NUCLEAR POWER

Analysis of Regional 

Differences and 

Improved Access to 

Information Could 

Strengthen NRC 

Oversight

Report to Congressional Requesters

September 2013

GAO-13-743

United States Government Accountability Office
United States Government Accountability Office

Highlights of GAO-13-743, a report to 

congressional requesters

September 2013

NUCLEAR POWER

Analysis of Regional Differences and Improved 

Access to Information Could Strengthen NRC

Oversight 

Why GAO Did This Study

The 2011 disaster at Japan's 

Fukushima Daiichi Nuclear Power 

Plant demonstrated that unexpected

nuclear accidents with extreme 

consequences can occur and, thus, 

heightened concerns about NRC’s 

ability to oversee the safety of U.S. 

commercial nuclear power reactors. 

NRC oversees safety through multiple 

processes, such as physically 

inspecting reactors and also 

responding to signs of declining 

performance (i.e., findings) or 

violations of its requirements. 

GAO was asked to review NRC’s 

oversight of the U.S. nuclear power 

industry. This report examines (1) how 

NRC implements its processes for 

overseeing the safety of commercial 

nuclear power reactors; (2) the extent 

to which NRC consistently identifies 

and resolves findings through these

processes; and (3) NRC’s methods for 

developing lessons learned to improve 

its oversight and challenges, if any, 

NRC faces in doing so. GAO reviewed 

NRC policies and guidance; visited five 

nuclear power plants located in 

multiple NRC regions; analyzed NRC 

data on findings, violations, licensee

performance, and inspection hours; 

interviewed NRC officials and industry 

representatives; and observed 

demonstrations of NRC’s database

search tools. 

What GAO Recommends

GAO recommends, among other 

things, that NRC analyze the causes of 

differences in identifying and resolving 

findings across regional offices and

address these differences, and that it

improve its database search tools. 

NRC agreed with GAO’s

recommendations.

What GAO Found

The Nuclear Regulatory Commission (NRC) relies on its staff’s professional 

judgment in implementing its processes for overseeing the safety of U.S. 

commercial nuclear power reactors. In implementing this oversight, NRC 

allocates specific roles and responsibilities to resident inspectors assigned to

each plant, regional officials at one of four regional offices responsible for most 

oversight activities, headquarters officials, and the nuclear power industry. NRC 

also builds into its processes incentives for plant managers to identify concerns 

about reactor safety, report those concerns to NRC, and take prompt actions to 

correct them. NRC’s processes for identifying and assessing findings and 

violations are based on prescribed agency procedures and include several points 

where NRC staff must exercise their professional judgment, such as determining 

whether issues of concern identified during physical inspections constitute 

findings or violations and the risk significance of any findings or the severity of 

any violations, among other things. 

NRC is aware of differences across regional offices in identifying and resolving 

findings that result from physical inspections. GAO’s analysis of NRC’s data 

indicated that the number of escalated findings had fewer differences across 

regions than nonescalated findings, which are lower-risk findings and less severe 

violations. According to NRC officials, several factors, such as the hours spent on 

inspections, may explain the differences in nonescalated findings. However, 

GAO found that the regional office with the fewest reactors and the fewest 

inspection hours had the most nonescalated findings. NRC officials and industry 

representatives have raised concerns that the differences may also be due to 

differences in how NRC staff identify and resolve findings. NRC has taken some 

steps to examine the consistency of its oversight. For example, in 2009, the four 

regional offices implemented an initiative to explore how the regional offices

identify and assess inspection findings. However, NRC has not conducted a 

comprehensive analysis of the causes of the differences in the number of 

nonescalated findings across regions. Under federal standards for internal 

control, managers are to compare actual performance with planned or expected

results throughout the organization and analyze significant differences. Without 

such an analysis, NRC does not know whether its regional offices are applying 

regulations and guidance consistently. 

NRC has both formal and informal methods for developing lessons learned to 

improve its oversight. Formal methods include agencywide programs, annual and 

biennial assessments, and special initiatives. Informal methods include reaching 

out to peers and technical experts across the agency and accessing various 

agency databases. Although NRC guidance directs inspectors to use information 

in agency databases on past experiences to plan and conduct inspection 

activities, inspectors face challenges accessing this information, which may limit 

their ability to use it. For example, several NRC inspectors reported contacting 

other inspectors informally because NRC’s database search tools contain limited 

instructions and do not ensure thorough results. Without better search tools, 

inspectors may overly rely on information available through informal channels. View GAO-13-743. For more information, 

contact Frank Rusco at (202) 512-3841 or 

ruscof@gao.gov.
Page i GAO-13-743 Oversight of Reactor Safety

Letter 1 

Background 5

NRC Relies on Staff’s Professional Judgment in Implementing Its 

Processes for Overseeing the Safety of Commercial Nuclear 

Reactors 16

Differences Exist across NRC Regions in Identifying and Resolving 

Findings, and NRC Has Taken Some Steps to Address Them 26

NRC Uses Numerous Methods to Develop Lessons Learned for 

Improving Oversight, but Challenges Remain in Accessing This 

Information 44

Conclusions 49

Recommendations for Executive Action 51

Agency Comments 51

Appendix I Objectives, Scope, and Methodology 53

Appendix II Summary Information about Operating Nuclear Power 

Reactors in the United States 56

Appendix III NRC Safety Cornerstones and Related Performance 

Indicators 60

Appendix IV NRC Safety Cornerstones and Related Inspection Procedures 63

Appendix V Examples of NRC Findings and Violations 69

Appendix VI NRC Nonescalated Findings at Operating Nuclear Power 

Plants in the United States 71

Contents 
Page ii GAO-13-743 Oversight of Reactor Safety

Appendix VII Summary of NRC Performance Assessments for Nuclear 

Power Reactors Operating from 2001 through 2012 74

Appendix VIII Substantive Cross-cutting Issues Identified by NRC at 

Operating Nuclear Power Reactors in the United States 81

Appendix IX NRC Escalated Findings at Operating Nuclear Power Plants

in the United States 86

Appendix X NRC Fukushima Near-Term Task Force Recommendations 90

Appendix XI Comments from the Nuclear Regulatory Commission 92

Appendix XII GAO Contact and Staff Acknowledgments 95

Tables

Table 1: Strategic Areas and Cornerstones of NRC’s Oversight 7

Table 2: NRC Categories of Performance Indicators 9

Table 3: NRC Categories of Risk Significance for Inspection 

Findings 11

Table 4: NRC Reactor Oversight Process Action Matrix 14

Table 5: Nonescalated Findings per NRC Region, Calendar Years 

2000-2012 27

Table 6: Nonescalated Findings and Inspection Hours per NRC 

Region, Calendar Years 2000-2012 29

Table 7: Nonescalated Findings and Amount of Time under 

Increased Oversight (per NRC’s Action Matrix), Calendar 

Years 2001-2012 30

Table 8: Amount of Time Reactors per NRC Region Have Been in 

Commercial Operation 31
Page iii GAO-13-743 Oversight of Reactor Safety

Table 9: Number of Reactors per NRC Region Associated with 

Licensee Fleets of Commercial Nuclear Power Plants 32

Table 10: Number of Findings Assigned at Least One Cross-cutting 

Aspect per NRC Region, Calendar Years 2000-2012 34

Table 11: Types of Cross-cutting Aspects Assigned to Findings per 

NRC Region, Calendar Years 2000-2012 35

Table 12: Number of Nuclear Power Plants Assigned Substantive 

Cross-cutting Issues by NRC Region, Calendar Years 

2001-2012 36

Table 13: Escalated Findings per NRC Region, Calendar Years 

2000-2012 37

Table 14: Summary Information about Operating Commercial 

Nuclear Power Reactors 56

Table 15: NRC Safety Cornerstones and Related Baseline 

Inspection Procedures 63

Table 16: NRC Supplemental Inspection Procedures 67

Table 17: NRC Reactive Inspection Procedures 67

Table 18: Select Examples of NRC Findings Illustrating Different 

Risk Significance Levels 69

Table 19: Select Examples of NRC Violations Illustrating Different 

Severity Levels 70

Table 20: NRC Nonescalated Findings (Green Inspection Findings 

and Severity Level IV Violations) per Operating 

Commercial Nuclear Power Plant, Calendar Years 2000-

2012 71

Table 21: Highest Level of Oversight Applied by NRC to Operating 

Commercial Nuclear Power Reactors for One or More 

Quarters of Each Calendar Year, 2001-2012 74

Table 22: Substantive Cross-cutting Issues Identified by NRC at 

Operating Commercial Nuclear Power Reactors for One or 

More Quarters of Each Calendar Year, 2001-2012 81

Table 23: NRC Escalated Findings (Greater-than-Green Inspection 

Findings and Severity Level I, II, and III Violations) per 

Operating Commercial Nuclear Power Plant, Calendar 

Years 2000-2012 86

Figures

Figure 1: NRC Regions and Operating Nuclear Power Plants in the 

United States 6

Figure 2: NRC’s Processes for Identifying and Assessing Findings 

and Violations at Nuclear Power Reactors 25
Page iv GAO-13-743 Oversight of Reactor Safety

Figure 3: Nonescalated Findings by NRC Region and by Year, 

Calendar Years 2000-2012 28

Abbreviations

ADAMS Agencywide Documents Access and Management System

ALARA as low as reasonably achievable

IG Office of the Inspector General

INPO Institute of Nuclear Power Operations

NFPA National Fire Protection Association

NRC Nuclear Regulatory Commission

ROE Reactor Operating Experience

ROP Reactor Oversight Process

SERP Significance and Enforcement Review Panel

This is a work of the U.S. government and is not subject to copyright protection in the 

United States. The published product may be reproduced and distributed in its entirety 

without further permission from GAO. However, because this work may contain 

copyrighted images or other material, permission from the copyright holder may be 

necessary if you wish to reproduce this material separately.
Page 1 GAO-13-743 Oversight of Reactor Safety

441 G St. N.W.

Washington, DC 20548

September 27, 2013

The Honorable Barbara Boxer

Chairman

Committee on Environment and Public Works

United States Senate

The Honorable Sheldon Whitehouse

Chairman

Subcommittee on Oversight

Committee on Environment and Public Works

United States Senate

The Honorable Robert Menendez

United States Senate

The Honorable Bernard Sanders

United States Senate

In March 2011, an earthquake off the Pacific coast of Japan triggered a 

tsunami that caused significant accidents at three reactors in the 

Fukushima Daiichi Nuclear Power Plant complex, demonstrating that lowprobability, high-consequence nuclear accidents can and do occur.1

According to the Japanese parliament’s formal investigation, the damage 

to the Fukushima reactors could have been avoided and resulted, in part, 

from a failure of regulatory oversight in Japan. This event heightened 

concerns about the safety of commercial nuclear power plants worldwide 

and the ability of the Nuclear Regulatory Commission (NRC)—the 

independent federal agency responsible for licensing and regulating U.S. 

civilian nuclear activities—to oversee the safe operation of commercial 

nuclear power reactors in the United States.2 In response to these 

concerns, NRC convened a Fukushima Near-Term Task Force, which in 

a July 2011 report concluded that a similar accident is unlikely to occur in 

 

1

Low-probability, high-consequence events are accidents that are unexpected, with few 

similar historical events; however, when they do happen, the impact is extreme. Such 

events include natural disasters, terrorism, nuclear power plant accidents, and chemical 

plant explosions. 

2

In this report, when we use the term nuclear power plant, we are referring to an entire 

site that uses one or more nuclear power reactors to generate electricity. 

 
 

Page 2 GAO-13-743 Oversight of Reactor Safety

the United States and identified many opportunities for improving NRC’s 

oversight.3

NRC’s mission is, in part, to ensure adequate protection of public health 

and safety from accidents involving commercial nuclear power reactors. 

In fiscal year 2012, the licensing and regulating of commercial reactors 

comprised over 75 percent—or approximately $800 million—of NRC’s 

budget, of which the vast majority (87 percent) is collected in the form of 

fees from industry. One of NRC’s stated goals is to ensure that its 

oversight of reactor safety is objective, predictable, and understandable, 

and according to NRC’s strategic plan, a key strategy in achieving its 

mission is to conduct this oversight openly to inform the public and 

maintain confidence in reactor safety. However, NRC’s Office of the 

Inspector General (IG) reported in 2010, 2011, and 2012, that one of the 

most serious management and performance challenges facing NRC is 

the ability to effectively carry out its oversight responsibilities while 

responding to emerging demands associated with licensing and 

regulating new reactors.4

NRC’s oversight will soon likely take on even greater importance as many 

commercial reactors in the United States are reaching or have reached 

the end of their initial 40-year operating period.5 Nuclear power’s 

contribution to the nation’s supply of electricity—about 20 percent in 

2012—will depend, in large part, on the continued operation of the 

nation’s 100 operating commercial reactors. Most of these reactors have

 

3

NRC, Recommendations for Enhancing Reactor Safety in the 21st Century: The NearTerm Task Force Review of Insights from the Fukushima Dai-ichi Accident (July 12, 

2011). 

4

NRC, Office of the Inspector General, Evaluation Report: Inspector General’s 

Assessment of the Most Serious Management and Performance Challenges Facing NRC, 

OIG-13-A-01 (Oct. 1, 2012); Evaluation Report: Inspector General’s Assessment of the 

Most Serious Management and Performance Challenges Facing NRC, OIG-12-A-01 (Oct. 

3, 2011); and Evaluation Report: Inspector General’s Assessment of the Most Serious 

Management and Performance Challenges Facing NRC, OIG-11-A-01 (Oct. 1, 2010). 

5

NRC issues licenses for commercial nuclear power reactors to operate for up to 40 years 

and allows these licenses to be renewed for up to an additional 20 years. There is no set 

limit on the number of times a reactor’s operating license can be renewed. 
 

Page 3 GAO-13-743 Oversight of Reactor Safety

had their licenses renewed to operate for another 20 years.6 However, as 

we recently reported,7 NRC’s license renewal process primarily considers 

issues related to the negative effects of aging on reactors and their 

components. Other ongoing safety matters are not explicitly covered in 

the license renewal process because NRC considers its ongoing 

oversight processes as adequate to ensure that reactors are operated 

safely and will continue to be in the future.

We and others have reported on NRC’s oversight of reactor safety. In 

September 2006, we reported that NRC had taken several steps to 

improve its oversight of nuclear reactors but had been slow to act on 

needed improvements, in particular, in improving the agency’s ability to 

identify and address early indications of declining safety performance.8 In 

2008, NRC’s IG found that NRC’s regional offices were not consistently 

implementing the agency’s enforcement program.9 Additionally, the 

National Academy of Sciences is conducting a study and preparing a 

report to NRC and Congress on lessons learned from the Fukushima 

nuclear accident for improving the safety and security of U.S. nuclear 

power plants. According to the National Academy of Sciences, this report 

is expected to be issued publicly by the summer of 2014.

 

6

Most of the nation’s operating commercial nuclear power reactors received their initial 

operating licenses in the 1970s and 1980s. By the end of 2013, 62 reactors will have held 

an operating license for at least 30 years, and of these reactors, 20 will have held an 

operating license for 40 years or more. As a result, many reactors are reaching or have 

reached the end of their initial 40-year operating period. As of May 2013, NRC had 

renewed 73 reactor licenses and was reviewing license renewal applications for 14 

reactors. However, in May 2013, the owner of the Kewaunee Power Station in Wisconsin 

permanently shut down that site’s reactor, and in June 2013, the owner of the San Onofre 

Nuclear Generating Station in California permanently shut down that site’s two reactors. 

The Kewaunee reactor was one of the 73 reactors with an operating license renewed by 

NRC, but the San Onofre reactors were not. As a result, as of July 2013, there are 72 

commercial nuclear power reactors with renewed operating licenses operating in the 

United States. 

7

GAO, Nuclear Reactor License Renewal: NRC Generally Follows Documented 

Procedures, but Its Revisions to Environmental Review Guidance Have Not Been Timely, 

GAO-13-493 (Washington, D.C.: May 30, 2013). 

8

GAO, Nuclear Regulatory Commission: Oversight of Nuclear Power Plant Safety Has 

Improved, but Refinements Are Needed, GAO-06-1029 (Washington, D.C.: Sept. 27, 

2006). 

9

NRC, Office of the Inspector General, Audit of NRC’s Enforcement Program, OIG-08-A17 (Sept. 26, 2008). 
 

Page 4 GAO-13-743 Oversight of Reactor Safety

In this context, you asked us to review NRC’s oversight of the U.S. 

commercial nuclear power industry. This report: (1) describes how NRC 

implements its processes for overseeing the safety of commercial nuclear 

power reactors; (2) evaluates the extent to which NRC consistently 

identifies and resolves findings through these processes; and (3) 

describes NRC’s methods for developing lessons learned to improve its 

oversight and challenges, if any, NRC faces in doing so.

To address these objectives, we reviewed relevant federal laws, 

regulations, and agency guidance covering NRC’s processes for 

overseeing reactor safety. In addition, we reviewed previous GAO and 

NRC IG reports that identified safety-related challenges or made 

recommendations to improve NRC oversight. We also reviewed NRC’s 

efforts to update its oversight processes. We interviewed NRC 

headquarters officials, regional officials, and inspectors. We also visited a 

nonprobability sample of five selected nuclear power plants and 

conducted interviews with the NRC inspectors and plant managers and 

operators at the plants.10 To select the five plants we visited—Calvert 

Cliffs Nuclear Power Plant in Maryland, Byron Station in Illinois, Palisades 

Nuclear Plant in Michigan, Cooper Nuclear Station in Nebraska, and Wolf 

Creek Generating Station in Kansas—we used multiple criteria, including 

whether a plant’s performance (as determined by NRC) has changed 

(i.e., by improving or declining) in the past 2 years, and whether, 

collectively, plants were located in multiple NRC regions, operated by 

different companies that hold licenses to operate reactors (i.e., licensees), 

and included different reactor types. We also analyzed NRC performance 

data for all 65 plant sites from 2000 through 2012. To assess the 

reliability of NRC’s data, we reviewed existing information about the data 

and the system that produced them and interviewed agency officials 

knowledgeable about the data. We found that these data were sufficiently 

reliable for the purposes of our report. Our review focuses on NRC’s 

oversight of the safe operation of reactors; physical security of plants is 

outside the scope of this review. A detailed description of our objectives, 

scope, and methodology is presented in appendix I.

 

10Because this was a nonprobability sample, the information we gathered from these visits 

to these plants is not generalizable to all 65 U.S. nuclear power plant sites but provides 

important illustrative information for understanding the oversight of reactors at the selected 

plants. 
 

Page 5 GAO-13-743 Oversight of Reactor Safety

We conducted this performance audit from May 2012 to September 2013 

in accordance with generally accepted government auditing standards. 

Those standards require that we plan and perform the audit to obtain 

sufficient, appropriate evidence to provide a reasonable basis for our 

findings and conclusions based on our audit objectives. We believe that 

the evidence obtained provides a reasonable basis for our findings and 

conclusions based on our audit objectives.

This background section describes NRC’s role and its processes for 

overseeing reactor safety, specifically the Reactor Oversight Process and 

the Traditional Enforcement Process.

NRC is an independent agency of about 4,000 employees established by 

the Energy Reorganization Act of 1974 to license and regulate U.S. 

civilian—that is, commercial, industrial, academic, and medical—use of 

nuclear materials. NRC is headed by a five-member Commission. The 

President appoints these Commission members, who are confirmed by 

the Senate, and designates one of them to serve as the Chair and official 

spokesperson. NRC seeks to carry out its mission of ensuring safe 

operation of the 100 commercial nuclear power reactors currently 

operating in the United States from its headquarters office in Rockville, 

Maryland, its four regional offices—in King of Prussia, Pennsylvania 

(Region I); Atlanta, Georgia (Region II); Lisle, Illinois (Region III); and 

Arlington, Texas (Region IV)—and resident (on-site) inspectors’ offices 

located at 62 nuclear power plants in 31 states.11 (See fig. 1.) Appendix II 

provides a complete list of the commercial nuclear power reactors

operating in the United States and additional information about each one.

 

11In this report, we use “operating” to describe those commercial nuclear power reactors 

and plants currently licensed by NRC to operate in the United States, regardless of 

whether they may be temporarily shut down (outages) for refueling, maintenance, 

inspection, emergency, or other nonpermanent purposes. Licensees must submit formal 

notification to NRC to permanently shut down a reactor. In February 2013, the owner of 

the Crystal River Nuclear Plant in Florida permanently shut down that plant’s reactor; in 

May 2013, the owner of the Kewaunee Power Station in Wisconsin permanently shut 

down that plant’s reactor; and in June 2013, the owner of the San Onofre Nuclear 

Generating Station in California permanently shut down that plant’s two reactors. These 

actions reduced the number of operating commercial nuclear reactors in the United States 

from 104 to 100, and they reduced the number of operating nuclear power plants from 65 

to 62. 

Background 

Role of NRC
 

Page 6 GAO-13-743 Oversight of Reactor Safety

Figure 1: NRC Regions and Operating Nuclear Power Plants in the United States

Under the Atomic Energy Act, as amended, NRC is responsible for 

issuing licenses to commercial nuclear reactors and conducting oversight 

of activities under such licenses to protect the health and safety of the 

public, among other things. NRC is authorized to conduct inspections and 

investigations, to enforce its requirements, such as by issuing orders and 

imposing civil (monetary) penalties, and to revoke licenses. To oversee 

the activities of licensed commercial reactors, NRC has established 

regulations, as well as guidance, including regulatory guides, standard 

review plans, and NRC’s Inspection Manual. These regulations and 

guidance, along with plant-specific licenses and technical specifications, 

form the basis by which NRC provides continuous oversight of reactor 

operations.
 

Page 7 GAO-13-743 Oversight of Reactor Safety

NRC’s oversight of commercial nuclear reactors includes monitoring their 

performance in three strategic areas: (1) reactor safety, that is, avoiding 

accidents and reducing the consequences of accidents if they occur; (2) 

radiation safety, that is, ensuring safety from radiation for both plant 

workers and the public during routine operations; and (3) safeguards, that 

is, protecting of the plant against sabotage or other security threats. 

These areas comprise seven cornerstones of NRC’s oversight (see table 

1). According to NRC, reactors’ satisfactory performance in these 

cornerstones provides reasonable assurance of safe operation. Our 

review focuses on NRC’s oversight of the safe operation of reactors; 

therefore, the physical security of plants (Physical Protection cornerstone) 

is outside the scope of this review.

Table 1: Strategic Areas and Cornerstones of NRC’s Oversight 

Strategic 

area Cornerstone Description

Reactor 

safety 

Initiating Events This cornerstone focuses on operations and events at a nuclear plant that could lead to a 

possible accident, if plant safety systems did not intervene. These events could include 

equipment failures leading to a plant shutdown, shutdowns with unexpected complications, 

or large changes in the plant’s power output. 

Mitigating Systems This cornerstone measures the function of safety systems designed to prevent an accident 

or reduce the consequences of a possible accident. The equipment is checked by periodic 

testing and through actual performance. 

Barrier Integrity This cornerstone focuses on the physical barriers between the highly radioactive materials 

in fuel within the reactor and the public and the environment outside the plant. These 

barriers are the sealed rods containing the fuel pellets, the heavy steel reactor vessel and 

associated piping, and the reinforced concrete containment building surrounding the 

reactor. The integrity of the fuel rods, the vessel, and the piping is continuously checked for 

leakage, while the ability of the containment to prevent leakage is measured on a regular 

basis. 

Emergency 

Preparedness

This cornerstone measures the effectiveness of plant staff in carrying out emergency plans 

for responding to a possible accident. Such emergency plans are tested every two years 

during emergency exercises involving the plant staff and local, state, and in some cases, 

federal agencies. 

Radiation 

safety 

Occupational Radiation 

Safety

NRC regulations set a limit on radiation doses received by plant workers, and this 

cornerstone monitors the effectiveness of the plant’s program to control and minimize 

those doses. 

Public Radiation Safety This cornerstone measures the procedures and systems designed to minimize radioactive 

releases from a nuclear plant during normal operations and to keep those releases within 

federal limits. 

Safeguards Physical Protection Nuclear plants are required to have well-trained security personnel and a variety of 

protective systems to guard vital plant equipment, as well as programs to assure that 

employees are constantly fit for duty through drug and alcohol testing. This cornerstone 

measures the effectiveness of security and fitness-for-duty programs.

Source: GAO analysis of NRC information.
 

Page 8 GAO-13-743 Oversight of Reactor Safety

NRC also looks for systemic problems by monitoring reactors’ 

performance in three cross-cutting areas that extend across the 

cornerstones:

• Problem identification and resolution. The licensee (1) ensures that 

issues potentially affecting nuclear safety are promptly identified, fully 

evaluated, and addressed in a timely manner; (2) uses operating 

experience information to support plant safety; and (3) conducts 

performance assessments to identify areas for improvement.

• Human performance. The licensee (1) makes decisions 

demonstrating that nuclear safety is an overriding priority; (2) ensures 

that staff, equipment, procedures, and other resources are available 

and adequate to assure nuclear safety; (3) plans and coordinates 

work activities consistent with nuclear safety; and (4) uses work 

practices that support human performance.

• Safety-conscious work environment. The licensee maintains an 

environment in which employees feel free to raise safety concerns, 

both to their management and to NRC, without fear of retaliation and 

where such concerns are promptly reviewed, properly prioritized, and 

appropriately resolved.

The Reactor Oversight Process (ROP) is the central component of NRC’s 

oversight efforts. Under the ROP, NRC undertakes four key actions: (1) 

reviewing key indicators of licensees’ performance; (2) physically 

inspecting reactors; (3) assessing whether any findings resulting from 

inspections are associated with any of the cross-cutting areas; and (4) 

determining what actions, if any, to take in response to signs of declining 

performance and any related violations of NRC requirements.

NRC technical specialists located at its regional offices review licensee 

performance using quantitative measures or indicators of that 

performance. According to NRC documents, performance indicators are 

designed to illustrate a long-term view of a licensee’s performance—over 

multiple quarters. Currently, NRC uses 17 performance indicators it 

developed collaboratively with the industry to measure licensee 

performance. Each performance indicator is designed to help gauge a 

licensee’s performance within the different cornerstones outlined above. 

For example, a performance indicator under the “Initiating Events” 

cornerstone measures the number of unplanned reactor shutdowns 

during the previous four quarters. Appendix III provides a list of the 

The Reactor Oversight 

Process

Reviews of Performance 

Indicators
 

Page 9 GAO-13-743 Oversight of Reactor Safety

safety-related (nonsecurity) performance indicators and related 

cornerstones.

Each licensee voluntarily submits data quarterly for each performance 

indicator,12 and NRC measures licensee performance from these data 

against established numerical thresholds; assigning a color—green, 

white, yellow, or red—to reflect increasing risk to reactor safety based on 

these performance data. Specifically, a green indicator reflects 

performance within an expected range, but white, yellow, and red 

represent decreasing levels of a licensee’s performance (see table 2).

Table 2: NRC Categories of Performance Indicators 

Performance

indicator category Description

Green Indicates performance within an expected performance level where the associated cornerstone 

objectives are met.

White Represents performance outside an expected range of nominal utility performance, but related 

cornerstone objectives are still being met.

Yellow Indicates related cornerstone objectives are being met, but with a moderate reduction in the safety 

margin.

Red Signals a significant reduction in safety margin in the area measured by the performance indicator.

Source: NRC.

Physical inspections are one of the main tools NRC uses to oversee 

licensee performance. During these inspections, NRC inspectors are to 

verify the accuracy of the quarterly data for each performance indicator 

supplied by the licensee and to assess aspects of licensee performance 

that are not directly measured by the performance indicator data—for 

example, confirming that fire protection equipment is in place and that 

certain maintenance operations have occurred. These inspections cover 

a wide variety of major systems and technical areas of nuclear power 

reactors that NRC considers most critical to meeting the overall agency 

mission of ensuring nuclear power plant safety. NRC currently has 52 

different inspection procedures, and they are divided into three broad 

 

12According to NRC documents, licensees voluntarily submit performance indicator data 

electronically to NRC on a per-reactor basis under a program of the Nuclear Energy 

Institute (the commercial nuclear trade organization). The Nuclear Energy Institute 

develops—with input from NRC staff—and publishes the guidance for licensees to collect 

and report performance indicator data. According to NRC officials, NRC formally endorses 

the Nuclear Energy Institute’s proposals. 

Physical Inspections 
 

Page 10 GAO-13-743 Oversight of Reactor Safety

types of inspections—baseline (46), supplemental (3), and reactive (3)—

that vary by depth and objective. Appendix IV provides a list of inspection 

procedures.

Baseline inspections. Baseline inspections are the minimum level of 

inspections performed at all nuclear power reactors, regardless of 

performance, and are designed to detect declining safety performance in 

each of the cornerstones and to review licensee effectiveness at 

identifying and resolving safety problems. NRC resident inspectors, who 

are located at each plant, and regional technical specialists, who travel to 

each plant from NRC regional offices, conduct 46 baseline inspection 

procedures both as needed and at intervals ranging from quarterly to 

triennially, involving physical observations of licensee activities and plant 

performance, reviewing and verifying licensee reports, and interviewing 

licensee personnel. Each of the baseline procedures specifies a range of 

sample activities to inspect. Within the scope and requirements of each 

procedure, inspectors select the type and number of activities to review 

on the basis of factors, such as the sample activities available; their risk 

significance; and the amount of time since a particular system or 

component was last inspected.13 NRC assesses any issues of concern 

and determines whether to elevate them to a finding based on certain 

criteria. To determine whether an issue of concern constitutes a finding, 

inspectors must first decide whether it is associated with a performance 

deficiency, which NRC guidance defines as “an issue that is the result of 

a licensee not meeting a requirement or standard where the cause was 

reasonably within the licensee’s ability to foresee and correct, and 

therefore should have been prevented.” NRC guidance defines a finding 

as an issue of concern associated with a performance deficiency that 

inspectors or regional officials then determine to be more than minor.14

NRC then assesses—through its ROP Significance Determination 

 

13These inspections are focused on activities and systems that, if they fail, have the 

potential to trigger—or increase the consequences of—an accident, along with activities 

and systems that could mitigate the effects of such an accident. 

14“Minor issues” are defined by NRC as those that have little actual safety consequences, 

little or no potential to impact safety, little impact on the regulatory process, and no 

willfulness. For example, if a licensee missed providing an hourly update to a state agency 

during an event that resulted in no actual safety consequences, it would be considered 

minor if it did not detract significantly from the state agency’s ability to function during the 

emergency. 
 

Page 11 GAO-13-743 Oversight of Reactor Safety

Process—the finding’s risk significance.15 Similar to performance 

indicators, NRC assigns the finding one of four colors to reflect increasing 

risk to reactor safety. Green findings—also called nonescalated 

findings—equate to very low risk significance, while white, yellow, and red 

findings—considered to be escalated findings—represent increasing 

levels of risk (see table 3). Appendix V provides select examples of NRC 

findings and related information that illustrate the varying levels of risk 

significance.

Table 3: NRC Categories of Risk Significance for Inspection Findings

Inspection 

finding 

category Risk significance Description

Green Very low safety 

significance

Indicates that licensee performance is acceptable and cornerstone objectives are fully met 

with nominal risk and deviation.

White Low to moderate safety 

significance

Indicates that licensee performance is acceptable, but outside the nominal risk range. 

Cornerstone objectives are met with minimal reduction in safety margin.

Yellow Substantial safety 

significance

Indicates a decline in licensee performance that is still acceptable with cornerstone 

objectives met, but with significant reduction in safety margin.

Red High safety significance Indicates a decline in licensee performance that is associated with an unacceptable loss 

of safety margin. Sufficient safety margin still exists to prevent undue risk to public health 

and safety.

Source: NRC.

Supplemental inspections. When NRC identifies one or more escalated 

findings or a performance indicator exceeding the green threshold at a 

reactor, regional officials conduct supplemental inspections, which 

expand the scope beyond baseline inspection procedures and focus on 

diagnosing the cause of the finding. There are three levels of 

supplemental inspections that increase in breadth and depth of analysis, 

depending on the number and type of performance problems identified. 

The first and lowest level of supplemental inspection assesses the 

licensee’s corrective actions to ensure they were sufficient in both 

correcting the problem and addressing the root and contributing causes to 

 

15The measures used for the characterization of risk are risk of damage to the reactor 

core and risk of release of radiation. In some situations, risk calculations cannot be made 

using these measures, such as in the case of measuring the risk for emergency 

preparedness inspection findings. In these cases, thresholds were determined by panels 

of experts on the basis of operating experience and a determination of the appropriate 

response. 
 

Page 12 GAO-13-743 Oversight of Reactor Safety

prevent recurrence. The second level of supplemental inspection has an 

increased scope that includes independently assessing the extent of the 

condition for both the specific and any broader performance problems. 

The third and highest level of supplemental inspection is more 

comprehensive and includes determining whether the reactor can 

continue to operate and whether additional regulatory actions are 

necessary. The highest level of supplemental inspection is usually 

conducted by a multidisciplinary team of NRC inspectors and may take 

place over several months. Also, as a part of this inspection, NRC 

inspectors assess the adequacy of the licensee’s overall programs for 

identifying, evaluating, and correcting performance issues.

Reactive inspections. NRC conducts reactive inspections of licensee 

performance when specific events occur that are of particular interest to 

NRC because of their potential safety significance or because of potential 

generic safety concerns important to multiple licensees. Reactive 

inspections determine the cause of the event and assess the licensee’s 

response to the event. Reactive inspections have three levels that are 

normally determined by an event’s safety, or risk significance. According 

to NRC officials, recent reactive inspections have focused on issues such 

as an automatic reactor shutdown following a loss of offsite power, and a 

flaw discovered in a weld on a reactor component. To conduct reactive 

inspections, a team of experts is formed and an inspection charter issued 

that describes the scope of the inspection efforts.

NRC inspectors assess all of the findings that result from physical 

inspections to determine if any of the three cross-cutting areas 

contributed to a performance problem. If regional officials determine that 

one of the cross-cutting areas contributed to a finding, NRC assigns what 

it calls a cross-cutting aspect within that area to the finding. More than 

one cross-cutting aspect may be assigned to a finding, but this does not 

occur often, according to NRC officials. Every 6 months, NRC regional 

officials analyze all findings issued at each plant during the prior 12-

month assessment period. In general, if a nuclear plant has four findings 

with the same cross-cutting aspect assigned during the previous year, 

and NRC is concerned about the licensee’s progress in addressing them, 

NRC determines that the nuclear plant has what NRC calls a substantive 

cross-cutting issue.16 According to NRC documents, the agency identifies 

 

16NRC assigns substantive cross-cutting issues on a per-plant—rather than a perreactor—basis. 

Assessments of Findings for 

Cross-cutting Areas 
 

Page 13 GAO-13-743 Oversight of Reactor Safety

substantive cross-cutting issues to inform the licensee that it has a 

concern with its performance and to further encourage it to take 

appropriate actions before more significant performance issues emerge.

NRC assesses its reviews of performance indicators and all findings that 

result from its physical inspections to determine what actions, if any, to 

take in response to signs of declining performance, including violations of 

NRC requirements.17 The significance of any violations associated with a 

finding is incorporated into the color coded risk assessment made during 

the Significance Determination Process. For enforcement of violations 

associated with nonescalated findings, NRC establishes a public record 

of the violations and requires the licensee to take steps to correct the 

violations, but it does not usually require the licensee to document its 

responses. For enforcement of violations associated with escalated 

findings, NRC issues written notices of violation (to which licensees are 

typically required to respond), civil penalties, or orders.18

Based on NRC’s assessment of licensee performance under the ROP, 

NRC places each of the licensee’s reactors into one of five performance 

categories on its action matrix, which corresponds to graded, or 

increasing, levels of oversight. The action matrix is NRC’s formal method 

of determining how much additional oversight—mostly in the form of 

supplemental inspections—is required on the basis of the number and 

risk significance of performance indicators and inspection findings.19 NRC 

 

17In some cases, NRC may determine that a finding poses a safety risk even if it does not 

violate a regulatory or other requirement. In other cases, a licensee’s performance may 

violate a regulatory or other requirement, but NRC does not consider the issue to be a 

finding and therefore does not assign it a risk significance level (i.e., green, white, yellow, 

red) either because NRC has determined that the issue is a minor performance deficiency 

(i.e., less than very low safety significance), is not a performance deficiency (i.e., is not 

reasonably within the licensee’s ability to foresee and correct), or the safety risk cannot be 

quantified. 

18Orders can be issued to modify, suspend, or revoke a license; to cease and desist from 

a given practice or activity; or take such other action as may be proper. 

19NRC periodically removes plants from oversight under the action matrix—due to 

significant performance or operational concerns regarding plants that are shut down—

through a prescribed process that it refers to as the IMC 0350 process, which is named 

after the NRC Inspection Manual chapter providing guidance for this level of oversight. 

Upon implementation of this process, the appropriate regional administrator establishes 

an oversight panel—typically comprised of officials from both headquarters and regional 

offices, including senior management—that determines the oversight activities necessary 

to authorize the restart of the plant’s reactors. 

Response to Performance 

Indicators and Findings
 

Page 14 GAO-13-743 Oversight of Reactor Safety

considers licensee performance to be acceptable until it reaches column 

5 (see table 4).

Table 4: NRC Reactor Oversight Process Action Matrix 

Oversight 

category: 

Licensee 

response

(column 1)

Regulatory 

response

(column 2)

Degraded 

cornerstone

(column 3)

Multiple/repetitive 

degraded 

cornerstone

(column 4)

Unacceptable 

performance 

(column 5)

NRC’s 

assessment of 

licensee 

performance by 

reactor: 

All green findings 

and performance 

indicators

One white finding or 

performance 

indicator, or two 

white findings or 

performance 

indicators in 

different 

cornerstones

Two or more white

findings or 

performance 

indicators in one 

cornerstone, or one 

yellow finding or 

performance indicator, 

or any three white

findings or 

performance 

indicators

Two white findings or 

performance 

indicators or one 

yellow finding or 

performance indicator 

in one cornerstone for 

five or more quarters, 

or multiple yellow or 

one red finding or 

performance indicator

Overall unacceptable 

performance 

representing situations—

such as multiple 

significant violations of a 

plant’s license—in which 

NRC does not have 

reasonable assurance 

that the licensee can or 

will conduct its activities 

to ensure protection of 

public health and safety. 

In general, it is 

expected, but not 

required, that entry into 

column 4 will precede 

consideration of whether 

a plant is in column 5. 

NRC’s oversight 

actions: 

Baseline 

inspections only

Baseline inspections 

and first level of 

supplemental 

inspections

Baseline inspections 

and second level of 

supplemental 

inspections

Baseline inspections 

and third level of 

supplemental 

inspections

Order to modify, 

suspend, or revoke 

licensed activities a

NRC office 

responsible for 

oversight 

actions: 

Regional office Regional office Regional office Agency Agency b

Source: GAO analysis of NRC information.

b

Note: In addition to the actions listed in this table, increasingly higher levels of NRC management will 

meet with a licensee as it moves to the right on the action matrix.

a

For reactors at this oversight level, at a minimum, the licensee and NRC are to document agreement 

on the corrective actions the licensee will take through a performance improvement plan. NRC may 

also take actions including making a demand to the licensee for information or issuing an order up to 

and including a plant shutdown. 

b

Agency response involves senior management attention from both headquarters and regional 

offices.

On the basis of the results of the ROP, NRC provides licensees and the 

public with an overall assessment of each reactor’s performance. At the 

end of each 6-month period, NRC issues an assessment letter to each 

plant to describe what level of oversight the plant’s reactors will receive 

according to their placement on the action matrix, what actions NRC is 
 

Page 15 GAO-13-743 Oversight of Reactor Safety

expecting the licensee to take as a result of the performance issues 

identified, any specific enforcement actions NRC has taken, and any 

documented substantive cross-cutting issues. If a substantive crosscutting issue is identified, the letter will describe what actions NRC 

intends to take to monitor the issue. If the issue persists, NRC can 

request a number of actions from the licensee, including a written 

response describing the corrective actions the licensee intends to take. 

NRC also holds an annual public meeting at or near each plant to review 

its performance and address questions from members of the public and 

other interested stakeholders. Additionally, NRC has other mechanisms 

to make available its oversight results, such as a website that provides 

summaries of each reactor’s current overall performance.20 While it 

addresses most violations through the ROP, NRC also responds to some 

violations identified during the ROP under what it calls its Traditional 

Enforcement Process.21

NRC uses this process to assess and determine the enforcement actions, 

if any, to take in response to violations that, according to NRC guidance, 

cannot be addressed fully through the ROP. The process typically applies 

to violations that result in actual safety consequences, may impede 

NRC’s oversight of licensed activities, or involve deliberate misconduct or 

discrimination matters, as well as certain technical violations involving 

specific areas, such as operator licensing and fuel storage. The 

Traditional Enforcement Process, including possible civil penalties, is 

managed outside the ROP action matrix.22

 

20http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/actionmatrix_summary.html. Accessed 

July 15, 2013. 

21These are violations that resulted in actual safety or security consequences, including 

but not limited to violations resulting in radiation exposures to the public or plant personnel 

above regulatory limits; any violation during an actual general emergency that prevents 

offsite response organizations from implementing protective actions, under their 

emergency plans, to protect the public health or safety, and violations resulting in 

substantial releases of radioactive material; violations that may impact the ability of the 

NRC to perform its regulatory oversight function; violations involving willfulness; violations 

of NRC requirements for which there are no associated Significance Determination 

Process performance deficiencies. 

22All violations are subject to consideration for civil penalties, but according to NRC policy, 

violations assessed under the ROP—unlike those assessed under the Traditional 

Enforcement Process—are not typically considered for civil penalties unless they involve 

actual consequences. 

NRC’s Traditional 

Enforcement Process
 

Page 16 GAO-13-743 Oversight of Reactor Safety

The Traditional Enforcement Process focuses on a violation’s 

consequences and results in the assignment of a significance level 

ranging from Severity Level IV for the least significant to Severity Level I 

for the most significant concerns, according to NRC policy. NRC’s 

response to Severity Level IV violations is considered “nonescalated 

enforcement,” while its response to Severity Levels I, II, and III is 

considered “escalated enforcement.” In assigning a Severity Level to the 

violation, NRC assesses the following: (1) the actual safety 

consequences; (2) the potential safety consequences; (3) the potential for 

impacting NRC’s ability to perform its regulatory function (e.g., failure to 

provide complete and accurate information); and (4) any willful aspects of 

the violation. Severity Levels I and II violations generally involve actual or 

high potential consequences to public health and safety. Severity Level III 

violations are cause for significant concern, and Severity Level IV 

violations are less serious but are of more than minor concern. Appendix 

V provides select examples of violations that illustrate varying levels of 

severity. NRC includes results from the Traditional Enforcement Process 

in the overall semiannual assessment of each plant’s performance that it 

provides to plant licensees and the public.

NRC relies on its staff’s professional judgment in implementing its 

processes, which are largely prescribed in guidance, for overseeing the 

safety of commercial nuclear power reactors. In implementing its 

oversight programs, NRC allocates specific roles and responsibilities to 

resident inspectors, regional staff, headquarters officials, and the nuclear 

power industry. NRC also provides incentives for licensees to identify and 

report any concerns about reactor safety to NRC. NRC oversight relies on 

its staff’s professional judgment in applying regulations and guidance, 

such as determining: (1) whether issues of concern identified in 

inspections constitute findings, (2) the risk significance of any findings, 

and (3) whether these findings have cross-cutting aspects, among other 

things.

NRC Relies on Staff’s 

Professional 

Judgment in 

Implementing Its 

Processes for 

Overseeing the Safety 

of Commercial 

Nuclear Reactors
 

Page 17 GAO-13-743 Oversight of Reactor Safety

To implement its oversight, NRC allocates specific roles and

responsibilities to regional officials, resident inspectors, headquarters 

officials, and the nuclear power industry. According to NRC officials, 

doing so allows decisions to be made at the appropriate level within the 

agency, but it also necessitates open communication between the various 

offices to effectively share information and ensure policies and processes 

are implemented consistently across the agency.

Officials from NRC’s four regional offices are responsible for 

implementing the ROP and the Traditional Enforcement Process and 

establish—consistent with agency guidelines—the region’s procedures for 

assessing issues of concern and resolving most findings.23 According to 

NRC documents, the inspection program is intended to provide regional 

administrators flexibility with the planning and application of inspection 

resources to deal with risk-significant issues and problems. According to 

NRC officials, regional branch chiefs supervise small groups of resident 

inspectors, facilitate information sharing among inspectors in their region, 

and manage specialized teams of inspectors that conduct periodic plant 

inspections focused on specific technical areas, such as fire safety. 

Further, NRC officials said that regional branch chiefs are largely 

responsible for deciding whether an issue of concern raised during 

physical inspections constitutes a finding. Agency guidance provides 

them with relative autonomy in determining when an issue constitutes a 

more-than-minor performance deficiency.24

Since NRC established the ROP in 2000, regional offices have been 

responsible for addressing all nonescalated findings and violations, which 

constituted about 90 percent of findings and violations, according to our 

analysis of NRC data. In addition, though subject to oversight by NRC 

headquarters, regional offices play a lead role when assessing and 

responding to escalated findings or violations, including requiring 

licensees to provide a written response describing the reasons for a 

 

23According to NRC officials, each region has its own guidance for implementing NRC’s 

policies. 

24Issues determined to be minor performance deficiencies do not meet the criteria for an 

inspection finding. 

NRC Allocates Specific 

Oversight Roles and 

Responsibilities among 

Regional Officials, 

Resident Inspectors, 

Headquarters Officials, and 

the Nuclear Power 

Industry 

Regional Officials 
 

Page 18 GAO-13-743 Oversight of Reactor Safety

performance deficiency, actual and proposed corrective actions, and the 

expected date of compliance. After assessing licensee performance 

based, in part, on the findings or violations, regional offices then conduct 

supplemental inspections as appropriate, according to NRC guidance.

NRC assigns resident inspectors to each nuclear power plant to conduct 

baseline inspections.25 According to NRC guidance, resident inspectors 

are responsible for routinely conducting plant status reviews that include 

observing control room activities; attending licensee meetings; and 

conducting walk-downs, or inspections, of various plant areas. In addition 

to these duties, resident inspectors are responsible for ensuring that plant 

representatives implement corrective actions within a “reasonable period 

of time” in response to nonescalated findings.26

NRC headquarters officials play a key role in addressing escalated 

findings and violations, those assigned the highest categories of risk 

significance or severity levels. For example, NRC officials told us that, 

before findings are assessed as being escalated, NRC headquarters 

officials participate in meetings with regional officials and resident 

inspectors to ensure that the assessment is accurate and any proposed 

enforcement actions are justifiable. Moreover, according to NRC 

guidance, officials in NRC headquarters must approve all enforcement 

actions against licensees. Specifically, this includes the Director of NRC’s 

Office of Enforcement, and, in certain cases, such as when responding to 

Severity Level I violations, the Deputy Executive Director, after 

consultation with the Commission. These actions can include anything 

from monetary penalties, to orders to take specific actions, to shutting 

down a plant.

In addition to NRC, the commercial nuclear power industry also plays a 

key role to ensure that nuclear plants operate safely and comply with 

regulations. According to NRC policy, licensees have the primary 

responsibility for safely operating their plants, as well as in accordance 

with their licenses and NRC regulations. Each plant has many physical 

structures, systems, and components, and licensees have numerous 

 

25According to NRC officials, NRC rotates its resident inspectors among plants every 7 

years. 

26According to NRC’s enforcement policy, a “reasonable period of time” is commensurate 

with the severity of the violation. 

Resident Inspectors

Headquarters Officials 

Nuclear Power Industry
 

Page 19 GAO-13-743 Oversight of Reactor Safety

activities under way 24 hours a day to ensure that plants operate safely. 

These plants were designed according to a “defense-in-depth” philosophy 

revolving around redundant, diverse, and reliable safety systems. For 

example, two or more key safety components are in place so that if one 

fails, another one is present to serve as a backup. These include safety 

components such as safety injection pumps, sources of power, and 

physical barriers to contain radiation.

Furthermore, the Institute of Nuclear Power Operations (INPO)—the 

industry’s self-regulator—is a nonprofit organization formed in 1979 by 

the nuclear power industry in response to the accident at Three Mile 

Island.27 INPO has a key role in ensuring that plants operate safely by 

seeking to promote high levels of safety and reliability in the operation of 

nuclear power plants. According to its website, INPO sets performance 

objectives, criteria, and guidelines for the nuclear power industry, 

conducts evaluations of nuclear power plants, and communicates lessons 

learned and best practices throughout the nuclear power industry. 

According to INPO information, all organizations with direct responsibility 

to operate or construct commercial nuclear power plants in the United 

States have maintained continuous membership in INPO since its 

inception, and are primarily responsible for its funding. INPO provides a 

variety of services to the nuclear power industry, including personnel 

training for key positions at nuclear power plants, periodic evaluations 

and peer reviews of operating plants that are focused on plant safety and 

reliability, and assessments of plant performance across the industry to 

help licensees to understand issues that arise at other plants. According 

to INPO information, the organization has conducted nearly 1,200 plant 

evaluations, inspecting each plant approximately every 2 years, since 

1980. Also, according to an INPO official, INPO uses peer pressure to 

influence plants to improve, and in an industry presentation, its president 

and chief executive officer stated that its policy of providing licensees with 

confidential evaluation reports allows it to have more open and candid 

 

27A reactor at the Three Mile Island Nuclear Generating Station in southeastern 

Pennsylvania partially melted down on March 28, 1979. This was the most serious 

accident in U.S. commercial nuclear power plant operating history. In response to this 

accident, NRC implemented several changes to increase oversight of nuclear plants. 

These changes include expanding NRC’s resident inspector program—first authorized in 

1977—to have at least two inspectors live nearby and work exclusively at each plant in the 

United States to provide daily surveillance of licensee adherence to NRC regulations; and 

integrating NRC observations, findings, and conclusions about licensee performance and 

management effectiveness into periodic, public reports. 
 

Page 20 GAO-13-743 Oversight of Reactor Safety

discussions with plants about problems and areas for improvement. In 

addition, the industry’s collective insurance company for operational 

losses—Nuclear Electric Insurance Limited—uses INPO ratings as a 

factor in setting insurance premiums for insuring for costs associated with 

certain long-term interruptions of electricity supply and damages to plants, 

among other things.28 According to INPO and NRC documents, INPO and 

NRC communicate on topics of mutual interest related to improving the 

performance of commercial nuclear power plants.

NRC also builds into its processes incentives for licensees to identify 

concerns about reactor safety, discuss those concerns with NRC, and 

identify actions to correct them. According to NRC guidance, these 

concerns—known as licensee-identified findings—include any findings 

the licensee is already addressing through the plants’ respective 

corrective action programs. If NRC’s preliminary assessment of a 

licensee-identified finding indicates that the finding is green, NRC does 

not consider it when assessing licensee performance, does not consider it 

to count toward substantive cross-cutting issues, and does not apply 

additional oversight or require additional response from the licensee. 

According to NRC officials, licensee-identified findings demonstrate that 

the licensee is taking a proactive approach and can identify and correct 

problems before NRC identifies them; therefore NRC gives licensees 

credit for identifying potential problems during their own inspection 

processes. Regarding licensee-identified violations, according to NRC’s 

enforcement policy, NRC should normally give the licensee “credit” when 

assessing for civil penalties—that is, the violation should be issued with 

no civil penalty—for identifying the violation and taking prompt and 

comprehensive corrective actions. On the other hand, the policy states 

that NRC should issue a civil penalty, regardless of whether the violation 

is licensee-identified, if NRC officials determine that the corrective actions 

have not been prompt and comprehensive.

Conversely, according to its enforcement policy, NRC considers some 

findings and violations—whether identified by a licensee or an inspector—

 

28NRC, in implementing requirements of the Price-Anderson Act as amended, requires all 

licensees of nuclear power plants to have financial protection including liability insurance 

to ensure that funds would be available in the event of a nuclear accident at a U.S. 

nuclear power plant. American Nuclear Insurers—an insurance pool comprised of various 

insurance companies in the United States—provides liability insurance for nuclear power 

plants. 
 

Page 21 GAO-13-743 Oversight of Reactor Safety

to be “self-revealing.” For example, according to an agency official, a 

licensee may discover a pool of oil on the floor next to a diesel generator, 

indicating that something is wrong with the equipment. NRC may consider 

such a finding to be self-revealing, even if the licensee identified it. Under 

NRC’s enforcement policy, it may give the licensee credit for such 

findings under certain circumstances, such as licensee effort in 

discovering the root cause of the finding.

NRC’s oversight processes are based in regulation, and its specific 

processes for identifying and assessing findings and violations are largely 

prescribed in guidance, but rely on several key points where NRC staff 

must exercise their professional judgment, such as determining (1) 

whether an issue of concern constitutes a finding or violation, (2) a 

finding’s risk significance or a violation’s severity, and (3) whether findings 

or violations have cross-cutting aspects, among other things.

According to NRC documents and officials, NRC inspectors use their 

professional judgment and a combination of education, experience, and 

prescribed agency procedures to make initial assessments as to whether 

an issue of concern identified during an inspection constitutes a finding. 

NRC officials told us that performance deficiency was defined broadly to 

allow officials to use their professional judgment when assessing issues 

of concern. If NRC inspectors determine that there is a performance 

deficiency, they and regional staff then determine whether the 

performance deficiency is more than minor. In doing so, these officials 

have NRC-established questions and guidance and use their professional 

judgment in applying them. If inspectors and regional staff determine that 

a performance deficiency is not more than minor, then it is not considered 

a finding and the matter is not pursued any further under the ROP. To 

help in making these determinations, inspectors can request additional 

information from licensees and consult NRC inspectors assigned to other 

nuclear plants, according to NRC officials. If resident inspectors conclude 

they cannot adequately assess an issue of concern, they can seek 

assistance from their regional office or NRC headquarters—typically from 

technical specialists or branch chiefs who supervise resident inspectors at 

two to three plants—to help make a determination. As we previously 

reported,29 inspectors also select the type and number of activities to 

 

29GAO-06-1029. 

NRC Oversight Relies on 

Its Staff’s Professional 

Judgment in Applying 

Regulations and Guidance

Determining What Constitutes 

a Finding
 

Page 22 GAO-13-743 Oversight of Reactor Safety

review during inspections, in part, on the basis of an inspector’s 

professional judgment, which is based on information such as reviews of 

the licensee’s corrective action program, allegations, or interviews with 

plant employees.

Professional judgment also plays a key role in regional officials’ 

determining the risk significance of a finding and the severity of a 

violation. If NRC determines that a performance deficiency is more than 

minor, it is considered a finding, and NRC regional officials assess its 

significance using the Significance Determination Process, which also 

requires the staff to exercise their professional judgment, according to 

NRC officials. This process provides an initial screening to identify those 

nonescalated findings—that is, those findings that do not result in a 

significant increase in plant risk and thus need not be analyzed further. 

Remaining findings are subjected to additional reviews to determine their 

risk significance, using the next phase of the Significance Determination 

Process—the Significance and Enforcement Review Panel (SERP).30

According to NRC guidance, this more detailed assessment may involve 

NRC risk experts from the appropriate regional office or NRC 

headquarters and further review by the licensee’s plant staff. The final 

outcome of the ROP review—evaluating whether the finding is green, 

white, yellow, or red—is used to determine NRC actions that may be 

needed. For any escalated violations identified during the ROP that are to 

go through the Traditional Enforcement Process, the severity level of 

violations is determined using a combination of NRC’s enforcement 

policy, professional judgment, and information obtained from the licensee 

during a conference with them, if necessary. NRC officials initially 

determine enforcement actions during an enforcement panel, and then 

make a final determination during a meeting, or caucus, both of which 

loosely correspond to the SERP.

 

30According to NRC’s Inspection Manual, the SERP provides a management review of the 

preliminary significance characterization and basis of findings that are potentially white, 

yellow, red, or greater than green. When necessary, based on the results of a Regulatory 

Conference or written response provided by the licensee, the SERP provides the 

management review of the final significance characterization and the basis of findings that 

are white, yellow, or red. No official agency preliminary significance determination of 

white, yellow, red, or greater than green is to be made without a SERP review. During the 

SERP, panel members will discuss the merits of the finding and reach consensus on (1) 

the statement of deficient licensee performance on which the inspection finding is based; 

(2) the safety significance of the finding, including assignment of preliminary or final color; 

and (3) the apparent violation and the regulatory requirements that should be cited. 

Determining a Finding’s Risk 

Significance and the Severity of 

a Violation
 

Page 23 GAO-13-743 Oversight of Reactor Safety

Regarding violations, NRC’s enforcement policy states that, because the 

regulation of nuclear activities does not lend itself to a mechanistic 

treatment in many cases, judgment and discretion must be exercised in 

determining the severity levels of violations and the appropriate 

enforcement sanctions. According to this policy, the use of judgment and 

discretion includes the decision to issue a notice of violation or to propose 

or impose a civil penalty and the amount of the penalty, after considering 

the general principles of NRC’s enforcement policy and the significance of 

the violation, as well as the surrounding circumstances. According to its 

enforcement policy, NRC may choose to exercise discretion and either 

escalate or mitigate enforcement sanctions or otherwise refrain from 

taking enforcement action within the Commission’s statutory authority. 

According to this policy, NRC may exercise judgment and discretion to 

determine the severity levels of violations and the appropriate 

enforcement sanctions to be taken. For example, NRC may refrain from 

issuing a notice of violation or a proposed civil penalty for a Severity Level 

II, III, or IV violation that is identified after NRC has taken enforcement 

action, if the violation is identified by the licensee as part of the corrective 

action for the previous enforcement action and the violation has the same 

or similar root cause as the violation for which enforcement action was 

previously taken, among other things. 

In addition, according to NRC guidance, inspectors should also use their 

professional judgment when assessing whether a finding is associated 

with any cross-cutting aspects. For example, as we reported,31 in 

analyzing the failure of a valve to operate properly at one plant, NRC 

inspectors determined that plant employees had not followed the correct 

maintenance procedures; thus, the finding was associated with the 

human performance cross-cutting area.

Both opening and closing a substantive cross-cutting issue requires the 

use of professional judgment. According to NRC guidance, before 

opening a substantive cross-cutting issue, regional officials must 

determine, among other things, whether the licensee is already taking 

actions to address the issue and, if so, whether these actions are timely, 

are commensurate with its significance, and will be effective in 

substantially mitigating it. Once a substantive cross-cutting issue has 

been opened, the regional office that opened it is responsible for 

 

31GAO-06-1029. 

Determining Whether Findings 

Have Cross-cutting Aspects 
 

Page 24 GAO-13-743 Oversight of Reactor Safety

establishing the criteria for closing it. These criteria may include having 

fewer findings with the same cross-cutting area in a subsequent 

assessment period or the regional staff’s increased confidence in the 

licensee’s ability to address the substantive cross-cutting issue. The 

decision to close a substantive cross-cutting issue is made by regional 

staff, including the senior resident inspector, branch chief, and other 

regional management, but NRC headquarters officials told us they also 

participate.

NRC can require licensees to take several actions to close substantive 

cross-cutting issues. According to NRC guidance, after first issuing a 

letter identifying a substantive cross-cutting issue to a licensee, NRC staff 

may follow up on these issues in multiple ways, including semiannual 

evaluations or additional inspections. In the second consecutive 

assessment letter identifying the same substantive cross-cutting issue, 

the regional office may consider requesting the licensee provide a 

response at a public meeting or provide a written response or hold a 

separate meeting. In the third consecutive letter, according to NRC 

guidance, the regional office would typically request that the licensee 

perform an independent assessment of safety culture, although the 

regional office could decide that a safety culture assessment is not 

necessary if the licensee has made reasonable progress addressing the 

issue. If the substantive cross-cutting issue continues beyond the third 

letter, the regional office may consider additional actions not prescribed 

by the action matrix developed in consultation with the Director of NRC’s 

Office of Nuclear Reactor Regulation and its Executive Director of 

Operations.32

Figure 2 outlines NRC’s processes for identifying and assessing findings 

and violations and key points where professional judgment plays a role.

 

32According to NRC’s website, the Office of Nuclear Reactor Regulation is responsible for 

accomplishing key components of NRC’s nuclear reactor safety mission. As such, the 

office conducts a broad range of regulatory activities in the four primary program areas of 

rulemaking, licensing, oversight, and incident response for commercial nuclear power 

reactors, and test and research reactors to protect the public health, safety, and the 

environment. The office works with the NRC regions and other offices to accomplish its 

mission and contribute to the agency mission.
 

Page 25 GAO-13-743 Oversight of Reactor Safety

Figure 2: NRC’s Processes for Identifying and Assessing Findings and Violations at Nuclear Power Reactors
 

Page 26 GAO-13-743 Oversight of Reactor Safety

NRC is aware of differences across regions in identifying and resolving 

the findings that result from its oversight processes and has taken some 

steps to address them. In particular, the extent to which nonescalated 

findings, which equate to very low risk significance, have been identified 

differs across regions.33 Several factors may explain these differences, 

but NRC officials and industry representatives have raised concerns that 

they may be due, in part, to differences in how NRC staff identify and 

resolve findings. The four NRC regions also differed in the number of 

cross-cutting aspects they assigned to findings. NRC has taken some 

steps to address these differences but has not undertaken a 

comprehensive review of them. The number of escalated findings, which 

equate to greater risk significance, were more similar across regions.34

However, the information NRC makes publicly available about findings 

and how they are resolved is difficult to search and assess.

Some regions identify more nonescalated findings than others, even 

though they oversee fewer reactors. For example, as table 5 indicates, 

from 2000 through 2012, Region IV recorded more nonescalated findings 

than each of the other three regions, even though it oversaw the fewest 

number of reactors at the fewest number of plants. During the same 

period, Region II—the region with the greatest number of reactors and 

plants—recorded the fewest number of nonescalated findings. More 

specifically, Region IV recorded over 70 percent more nonescalated 

findings than Region II even though Region IV oversaw over one-third 

fewer reactors; resulting in an average number of nonescalated findings 

per reactor for Region IV to be over 2.6 times that of Region II.

 

33In our analysis, we use “nonescalated findings” to refer to combined NRC data for (1) 

green findings identified through the ROP (i.e., equating to very low risk significance) and 

(2) Severity Level IV violations identified through the Traditional Enforcement Process 

(i.e., violations that are less serious, but are of more than minor concern resulting in no or 

relatively inappreciable potential safety consequences). 

34In our analysis, we use “escalated findings” to refer to combined NRC data for (1) 

greater than green (i.e., white, yellow, and red) findings identified through the ROP and (2) 

escalated violations (i.e., Severity Levels III, II, and I) identified through the Traditional 

Enforcement Process. 

Differences Exist 

across NRC Regions 

in Identifying and 

Resolving Findings, 

and NRC Has Taken 

Some Steps to 

Address Them

Extent to Which 

Nonescalated Findings 

Have Been Identified 

Differs across NRC 

Regions 
 

Page 27 GAO-13-743 Oversight of Reactor Safety

Table 5: Nonescalated Findings per NRC Region, Calendar Years 2000-2012 

Nonescalated findings, plants, and reactors

Region I

(Northeast) 

Region II 

(Southeast) 

Region III 

(Midwest) 

Region IV

(West) 

Number of nonescalated findings 2,518 1,885 3,148 3,225

Number of plants 17 18 16 14

Number of reactors 26 33 24 21

Average number of nonescalated findings per reactor 96.8 57.1 131.2 153.6

Source: GAO analysis of NRC data.

Note: This table includes data on the 104 commercial nuclear reactors operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear reactors in the United 

States from 104 to 100.

As figure 3 shows, from 2000 through 2012, the number of nonescalated 

findings identified differed across NRC’s four regions, but Region IV 

recorded more nonescalated findings than any of the other three regions 

for 6 consecutive years—2007 through 2012. Appendix VI lists the 

number of nonescalated findings identified at each plant from 2000 

through 2012.
 

Page 28 GAO-13-743 Oversight of Reactor Safety

Figure 3: Nonescalated Findings by NRC Region and by Year, Calendar Years 

2000-2012

Note: Region I oversees the Northeast; Region II oversees the Southeast; Region III oversees the 

Midwest; and Region IV oversees the West.

NRC officials stated that various factors may contribute to differences in 

the number of nonescalated findings identified across regions, including 

differences in the (1) number of hours spent inspecting reactors; (2) 

amount of time reactors are under increased oversight as a result of 

performance deficiencies; (3) age of reactors; and (4) benefits from the 

combined resources of multiple plants owned or managed by one 

licensee.

Differences in inspection hours. NRC officials stated that differences in 

the number of hours regions spent inspecting reactors may have 

contributed to the differences in the number of nonescalated findings 

across regions, because more inspection hours would lead to more 

findings. However, according to 2000 through 2012 NRC data, while the 

total number of inspection hours has varied across the regions, the 

number of inspection hours has been fairly consistent across regions 

Various Factors May 

Contribute to Differences 

in the Number of Findings 

Identified 
 

Page 29 GAO-13-743 Oversight of Reactor Safety

when accounting for the number of reactors in each region. When 

comparing the number of nonescalated findings in each region to the 

number of inspection hours and accounting for the regional differences in 

the number of reactors, we found that the number of nonescalated 

findings per inspection hour varies widely. Specifically, Region II identified 

the least number of nonescalated findings per inspection hour per reactor, 

and Region IV identified the most, with more than 3.5 times as many 

nonescalated findings per 1,000 inspection hours per reactor as Region II 

(see table 6).

Table 6: Nonescalated Findings and Inspection Hours per NRC Region, Calendar Years 2000-2012

Nonescalated findings, inspection hours, and reactors

Region I 

(Northeast) 

Region II 

(Southeast) 

Region III 

(Midwest) 

Region IV 

(West) 

Number of nonescalated findings 2,518 1,885 3,148 3,225

Number of inspection hours (thousands) 562.8 532.2 505.3 396.1

Number of reactors 26 33 24 21

Number of inspection hours (thousands) per reactor 21.6 16.1 21.1 18.9

Number of nonescalated findings per 1,000 inspection hours 4.5 3.5 6.2 8.1

Average number of nonescalated 

findings per 1,000 inspection hours per reactor 0.17 0.11 0.26 0.39

Source: GAO analysis of NRC data.

Note: This table includes data on the 104 commercial nuclear reactors operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear reactors in the United 

States from 104 to 100.

Differences in the amount of time reactors are under increased oversight. 

NRC officials stated that differences in the amount of time reactors are 

under increased oversight as a result of performance deficiencies—

defined as being in column 2 or higher on NRC’s action matrix—may 

contribute to differences in the number of nonescalated findings across 

regions, because increased oversight results in more scrutiny and more 

findings. However, according to 2001 through 2012 NRC data, the 

number of quarters that reactors in each region were under increased 

oversight were fairly consistent across regions when accounting for the 

number of reactors in each region. When comparing the number of 

nonescalated findings in each region to the number of quarters that 

reactors in each region were under increased oversight, and accounting 

for the regional differences in the number of reactors, we found that the 

results varied widely. Specifically, Region II identified the least number of 
 

Page 30 GAO-13-743 Oversight of Reactor Safety

nonescalated findings per quarter under increased oversight, and Region

IV recorded the most, with more than 4 times as many nonescalated 

findings per quarter under increased oversight as Region II (see table 7). 

Furthermore, Region II had more reactors under increased oversight than 

Region IV for 36 of the 48 quarters we reviewed. Appendix VII 

summarizes NRC’s performance assessments for each reactor from 2001 

through 2012.

Table 7: Nonescalated Findings and Amount of Time under Increased Oversight (per NRC’s Action Matrix), Calendar Years 

2001-2012

Nonescalated findings, quarters

under increased oversight, and reactors

Region I 

(Northeast) 

Region II 

(Southeast) 

Region III 

(Midwest) 

Region IV 

(West) 

Number of nonescalated findings 2,315 1,778 2,982 3,059

Number of quarters reactors were under increased 

oversight 274 a 375 305 233

Number of reactors 26 33 24 21

Average number of nonescalated findings per number 

of quarters under increased oversight per reactor 0.32 0.14 0.41 0.63

Source: GAO analysis of NRC data.

Notes: Based on NRC’s assessment of licensee performance under the ROP, NRC places each of 

the licensee’s reactors into one of five performance categories on its action matrix, which 

corresponds to graded, or increasing levels of oversight. The action matrix is NRC’s formal method of 

determining how much additional oversight—mostly in the form of supplemental inspections—is 

required on the basis of the number and risk significance of performance indicators and inspection 

findings.

This table includes data on the 104 commercial nuclear reactors operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear reactors in the United 

States from 104 to 100.

According to NRC officials, the number of inspection hours allocated to supplemental inspections 

conducted under increased oversight can vary by a factor of 75.

a

The period of our review breaks down to 48 quarters. Not every reactor was under increased 

oversight for each quarter.

Differences in the age of reactors. NRC officials stated that differences in 

the age of reactors may contribute to the differences in the number of 

nonescalated findings across regions because nonescalated findings 

become more prevalent as reactors age. We analyzed the age of reactors 

across the regions and identified regional differences in their age. 

However, we found that reactors in Region II, which had the fewest 

nonescalated findings from 2000 through 2012, have been in commercial 

operation, on average, more years than, and have collectively been in 
 

Page 31 GAO-13-743 Oversight of Reactor Safety

commercial operation for almost twice as long as, reactors in Region IV, 

which had the most nonescalated findings over that period (see table 8).

Table 8: Amount of Time Reactors per NRC Region Have Been in Commercial Operation

Years in commercial operation and reactors

Region I 

(Northeast) 

Region II 

(Southeast) 

Region III 

(Midwest) 

Region IV 

(West) 

Average number of years a reactor has been in 

commercial operation

33.9 33.4 34.2 28.3

Number of years reactors have collectively been in 

commercial operation 882 1103 822 594

Number of reactors 26 33 24 21

Source: GAO analysis of NRC data.

Note: This table includes data on the 104 commercial nuclear reactors operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear reactors in the United 

States from 104 to 100.

Differences in resources available to plants. NRC officials stated that, 

when one licensee owns or manages multiple plants (collectively referred 

to as a licensee’s “fleet”), benefits derived from plants’ combined 

resources may contribute to the differences in the number of 

nonescalated findings across regions, because some regions have more 

plants in fleets than others. For example, according to NRC officials, 

licensees often require all plants that they manage or own to make 

improvements in response to findings made at only one of their plants, 

which could lead all the plants in this fleet to experience fewer findings 

over time. NRC officials told us that plants associated with fleets typically 

perform better because the licensee holds all plants to the same 

performance levels and provides resources and shares what would 

otherwise be proprietary information to make that happen. Some 

independent plants—those not part of a fleet—share resources by 

forming alliances with other plants, entering into management contracts 

with larger companies, or both. However, NRC officials stated that 

independent plants are limited in how much information they can share 

because some information is considered to be proprietary. We analyzed 

whether individual plants were associated with licensee fleets and found 

that being associated with a fleet may be a factor in the number of 

nonescalated findings a region identifies. For example, Region II, which 

had the fewest nonescalated findings from 2000 through 2012, had the 

highest percentage of its reactors located at plants associated with fleets. 

Conversely, Region IV, which had the most nonescalated findings during 
 

Page 32 GAO-13-743 Oversight of Reactor Safety

the same period, had the lowest percentage of its reactors located at 

plants associated with fleets (see table 9). However, because NRC has 

not conducted a comprehensive analysis of several potential factors, it is 

not clear to the extent this factor influences the differences in the number 

of findings across regions.

Table 9: Number of Reactors per NRC Region Associated with Licensee Fleets of Commercial Nuclear Power Plants

Region

Number of reactors

associated with a

licensee’s fleeta Total number of

reactors in region Percentage

I (Northeast) 16 26 62

II (Southeast) 31 33 94

III (Midwest) 17 24 71

IV (West) 5 21 24

Total 69 104 66

Source: GAO analysis of NRC data.

Note: This table includes data on the 104 commercial nuclear reactors operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear reactors in the United 

States from 104 to 100.

a

For this report, a “fleet” is defined as reactors across two or more plants that are owned or managed 

by the same licensee.

In addition, NRC officials stated that the quality of reactor maintenance 

may contribute to the regional differences in the number of nonescalated 

findings. For example, the resident inspectors at a plant with a high 

number of findings in recent years told us that this was due, in part, to 

plant managers not fully addressing maintenance items for many years to 

save on costs. However, we could not analyze this factor because, 

according to NRC officials, the data are not readily available.

NRC officials and industry representatives have raised concerns that 

differences in the number of findings NRC regions identify may be due, in 

part, to differences in how NRC staff identify and resolve findings. NRC 

officials, including some resident inspectors in Region IV, told us that they 

believe Region IV identifies nonescalated findings more aggressively than 

other regional offices and that some regional branch chiefs differently 

interpret the criteria for assessing performance deficiencies and what 

constitutes more-than-minor findings. Some NRC officials also noted that 

NRC Officials and Industry 

Representatives Have 

Raised Concerns about 

Differences in the Number 

of Findings Identified 

across NRC Regions
 

Page 33 GAO-13-743 Oversight of Reactor Safety

the regions have flexibility to combine issues into a single finding or to 

split them into multiple findings, which may also account for some of the 

regional variation in the number of findings identified. These officials told 

us that, in their view, this autonomy and discretion is a good and 

desirable part of the ROP. Nonetheless, in 2008, NRC’s IG reported that 

NRC was inconsistently resolving findings across its regional offices due, 

in part, to not having clear and comprehensive guidance needed to 

facilitate consistency.

NRC officials told us they survey industry annually as part of a ROP selfassessment process and have received feedback since the 

implementation of the ROP in 2000 about concerns from the appearance 

of differences in identifying and resolving findings across regions. They 

also stated that the Nuclear Energy Institute—the commercial nuclear 

trade organization—frequently raises these concerns with them. Plant 

managers we interviewed—several of whom have worked at plants 

across different regions—also raised concerns about these apparent 

differences. These plant managers told us that such differences could 

lead to several challenges, including focusing limited plant resources on a 

greater number of issues of very low risk significance (i.e., nonescalated 

findings).

NRC officials also told us that another concern that is sometimes raised 

about its findings by industry is that findings at different plants that appear 

to be similar sometimes result in different risk assessments. NRC officials 

provided the example of two findings at two different plants that appeared 

similar but that led to different assessments. Specifically, both findings 

dealt with problems with certain components of auxiliary feedwater pumps 

that rendered the pumps inoperable. Even though these findings 

appeared similar, they resulted in different risk significance levels—one 

green and one white—because they affected the safety of the two plants 

differently. NRC officials told us the process of assessing the risk 

significance of a finding is inherently complex and includes several 

factors, such as

• design of the plant, including availability of backup equipment and 

whether the site uses cooling towers or lakes or rivers for its ultimate 

source of cooling water;

• the effect that the finding had on the performance and availability of 

safety equipment;

• the amount of time the equipment was inoperable;

• criticality of the safety equipment associated with the finding; and
 

Page 34 GAO-13-743 Oversight of Reactor Safety

• historical reliability of the equipment in question for that plant design, 

among others.

Once NRC accounts for these factors, similar findings may result in 

different assessments because the risk that a finding poses is plantspecific, according to NRC officials.

The four NRC regions also differed in the number of cross-cutting aspects 

they assigned to findings and the percentage of findings assigned crosscutting aspects. For example, Regions III and IV had about twice as many 

findings assigned at least one cross-cutting aspect as Region II (see table 

10).

Table 10: Number of Findings Assigned at Least One Cross-cutting Aspect per NRC Region, Calendar Years 2000-2012 

Region

Cross-cutting aspect

Findings

not assigned 

Findings assigned

at least one Total

Percentage of findings 

assigned at least one 

I (Northeast) 1,044 1,666 2,710 61

II (Southeast) 1,047 1,046 2,093 50

III (Midwest) 1,446 1,936 3,382 57

IV (West) 1,147 2,192 3,339 66

Total 4,684 6,840 11,524 59

Source: GAO analysis of NRC data.

Notes: If NRC officials determine that one of the cross-cutting areas—human performance, safetyconscious work environment, and problem identification and resolution—contributed to a finding, NRC 

assigns a cross-cutting aspect within that area to the finding. More than one cross-cutting aspect may 

be assigned to a finding.

NRC data did not include information about cross-cutting aspects for 397 findings.

This table includes data on the 104 commercial nuclear reactors operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear reactors in the United 

States from 104 to 100.

Table 11 shows the types of cross-cutting aspects assigned to findings 

across regions.

NRC Regions Also Differed 

in the Cross-cutting 

Aspects They Assigned to 

Findings 
 

Page 35 GAO-13-743 Oversight of Reactor Safety

Table 11: Types of Cross-cutting Aspects Assigned to Findings per NRC Region, Calendar Years 2000-2012 

Region

Cross-cutting aspect

Total

Percentage

total 

Safety-conscious

work environment Human performance

Problem identification 

and resolution

I (Northeast) 3 878 835 1,716 25

II (Southeast) 4 637 415 1,056 15

III (Midwest) 3 1,254 693 1,950 28

IV (West) 4 1,389 856 2,249 32

Total 14 4,158 2,799 6,971 100

Source: GAO analysis of NRC data.

Notes: If NRC officials determine that one of the cross-cutting areas—human performance, safetyconscious work environment, and problem identification and resolution—contributed to a finding, NRC 

assigns a cross-cutting aspect within that area to the finding. More than one cross-cutting aspect may 

be assigned to a finding.

NRC data did not include information about cross-cutting aspects for 397 findings.

This table includes data on the 104 commercial nuclear reactors operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear reactors in the United 

States from 104 to 100.

According to NRC guidance, identifying cross-cutting aspects is NRC’s 

primary means of identifying trends. The guidance states that a crosscutting aspect is a performance characteristic of a finding that is the most 

significant causal factor of the performance deficiency. The guidance 

further states that the ROP was developed with the presumption that 

inspection findings with cross-cutting aspects would help them identify the 

plants with significant performance issues. Therefore, according to NRC 

officials, the greater the number of findings with cross-cutting aspects, the 

greater the chance that NRC could assign a substantive cross-cutting 

issue to the plant. Once NRC assigns a substantive cross-cutting issue to 

a plant, closing the issue requires additional efforts by the licensee and 

additional oversight from NRC. The number of plants assigned 

substantive cross-cutting issues at some time during calendar years 2001 

through 2012 varied across the regions (see table 12). In Region II, less 

than half the plants were assigned a substantive cross-cutting issue, 

whereas a substantial majority of plants in other regions were assigned 

substantive cross-cutting issues. Appendix VIII summarizes the 

substantive cross-cutting issues assigned to plants from 2001 through 

2012.
 

Page 36 GAO-13-743 Oversight of Reactor Safety

Table 12: Number of Nuclear Power Plants Assigned Substantive Cross-cutting Issues by NRC Region, Calendar Years 

2001-2012 

Region

Nuclear power plants

assigned substantive

cross-cutting issues

Total nuclear

power plants

Percentage

assigned substantive

cross-cutting issues

I (Northeast) 12 17 71

II (Southeast) 7 18 39

III (Midwest) 15 16 94

IV (West) 12 14 86

Total 46 65 71

Source: GAO analysis of NRC data.

Notes: Every 6 months, NRC regional officials analyze all findings issued at each plant during the 

prior 12-month assessment period. In general, if a nuclear plant has four findings with the same 

cross-cutting aspect assigned during the previous year, and NRC is concerned about the licensee’s 

progress in addressing them, NRC determines that the nuclear plant has what NRC calls a 

substantive cross-cutting issue.

This table includes data on the 65 commercial nuclear power plants operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear power plants in the 

United States from 65 to 62.

According to NRC officials, because the criteria for closing a substantive 

cross-cutting issue involve professional judgment, determining whether to 

close one is somewhat subjective. The relevant regional office considers 

whether to close a substantive cross-cutting issue in its semiannual or 

annual assessment letters to licensees. Industry officials told us that, 

because NRC only considers closing substantive cross-cutting issues in 

these 6-month increments, it is difficult to close a substantive crosscutting issue in a timely fashion.

The number of escalated findings had fewer differences across regional 

offices. As shown in table 13, each region had between 46 and 93 

escalated findings from 2000 through 2012. Region III had more Severity 

Level III violations than the other regions, but otherwise, the number of 

escalated findings was largely consistent. Appendix IX lists the number of 

escalated findings identified at each plant from 2000 through 2012.

Number of Escalated 

Findings Had Fewer 

Differences across NRC 

Regions 
 

Page 37 GAO-13-743 Oversight of Reactor Safety

Table 13: Escalated Findings per NRC Region, Calendar Years 2000-2012 

Escalated findings, plants, and reactors

Region I 

(Northeast) 

Region II 

(Southeast) 

Region III 

(Midwest) 

Region IV

(West) 

Number of escalated findings 54 64 93 46

Number of plants 17 18 16 14

Number of reactors 26 33 24 21

Average number of escalated findings per reactor 2.1 1.9 3.9 2.2

Source: GAO analysis of NRC data.

Notes: The greater number of escalated findings in Region III over the other regions is due to the 

identification of more than three times as many Severity Level III violations.

This table includes data on the 104 commercial nuclear reactors operating in the United States 

through the end of calendar year 2012. In February 2013, the owner of the Crystal River Nuclear 

Plant in Florida permanently shut down that plant’s reactor; in May 2013, the owner of the Kewaunee 

Power Station in Wisconsin permanently shut down that plant’s reactor; and in June 2013, the owner 

of the San Onofre Nuclear Generating Station in California permanently shut down that plant’s two 

reactors. These actions reduced the number of operating commercial nuclear reactors in the United 

States from 104 to 100.

NRC officials told us that escalated findings are identified and addressed 

more consistently than nonescalated findings because they receive a 

much higher level of review than do nonescalated findings. For example, 

according to NRC’s enforcement policy, regional offices—subject to 

oversight by NRC headquarters—typically require licensees to provide a 

written response to an escalated finding that describes the reasons for 

the identified deficiency, actual and proposed corrective actions, and the 

expected date of compliance. Regional offices then conduct supplemental 

inspections for most escalated findings. Supplemental inspections expand 

the scope of baseline inspections to review, in part (1) the extent of the 

problem, (2) the sufficiency of the licensee’s evaluation of the root cause 

of the problem, and (3) the licensee’s proposed corrective actions. In 

addition, NRC officials told us that headquarters officials participate in 

meetings with regional officials and inspectors to ensure that an escalated 

finding is accurately assessed and that any proposed enforcement 

actions are justifiable. Moreover, according to NRC guidance, the Director 

of NRC’s Office of Enforcement (or designee)—located in headquarters—

has to approve any enforcement actions against licensees in response to 

escalated findings.
 

Page 38 GAO-13-743 Oversight of Reactor Safety

NRC has taken some steps to reduce differences across the regions in 

the identification and resolution of nonescalated findings but has not 

conducted a comprehensive analysis of the differences to determine its 

consistency in handling findings or reasons for the differences.

A key effort NRC has undertaken to reduce differences in how the regions 

implement the ROP is the ROP Reliability Initiative, which NRC’s regional 

offices initiated in 2009. According to NRC officials, the initiative was 

intended to explore how the different regional offices identify and assess 

inspection findings. Some NRC officials said that the initiative examines 

issues related to consistency, such as how to control the subjectivity in 

making distinctions between minor and more-than-minor performance 

deficiencies and violations. Each region took the lead on one of the 

following specific efforts:

• Enhanced inspection resource sharing among regions: This effort was 

a pilot program for sharing inspection resources across regional 

offices. Inspectors from different regions and resident inspectors from 

other plants participated on several types of team inspections. This 

initiative was led by Region IV and began in 2009.

• Branch chief visits to other regions: This effort included each region 

designating one branch chief to visit each of the other regions to 

identify best practices and implementation differences across the 

regions. This effort was led by Region II and began in 2009.

• Periodic discussion of reliability topics: This effort included quarterly 

videoconferences among regional officials to share insights on the 

application of ROP inspection guidance and gain insight into different 

regional decision making. This effort was led by Region I and began in 

2009.

• ROP self-assessments of inspection report quality:

35 This effort 

included a team of staff from all regions that selected a cross-section 

of inspection reports from each region to assess their conformance 

with applicable guidance, identify key differences between regions, 

and identify any weaknesses. This effort was led by Region III and 

concluded in 2012.

 

35NRC requires inspectors to document details supporting each finding or violation in an 

inspection report. 

NRC Has Taken Some 

Steps to Reduce 

Differences across Regions 

Concerning Nonescalated 

Findings but Has Not 

Conducted a 

Comprehensive Analysis
 

Page 39 GAO-13-743 Oversight of Reactor Safety

The ROP Reliability Initiative has not resulted in a formal report, and 

according to NRC officials, the sample of activities reviewed was not large 

enough to draw agencywide conclusions. In late 2011, NRC’s regional 

offices decided to continue the ROP Reliability Initiative by applying the 

approach of the individual efforts to an in-depth review of a specific 

inspection area each fiscal year. For example, the first such review done 

in fiscal year 2012 resulted in an internal memorandum making 25 

recommendations to provide additional guidance and training on limited 

aspects of NRC’s efforts to inspect for weaknesses in licensees’ 

corrective action programs. NRC officials told us that the overall initiative 

is still under way and that ongoing activities are intended to improve 

regions’ consistency of actions adhering to the ROP guidance; however, 

agency officials said that, due to the ROP Enhancement Project (see 

below) and increased inspection activities following the Fukushima 

accident, the regional offices have not identified a specific inspection area 

to focus on for the fiscal year 2013 review.

NRC has also conducted other efforts aimed at improving the consistency 

of its oversight processes:

• Assessment of regional enforcement. According to NRC officials, in 

2011, NRC established an initiative aimed at improving consistency in 

how regional offices resolve inspection findings through enforcement 

activities. This initiative was conducted to address findings of 

enforcement inconsistency in the 2008 report by the NRC IG.36

Officials stated that NRC recently completed four regional 

assessments that were intended to improve consistency of 

interpreting minor and more-than-minor thresholds for violations. 

According to NRC officials, this initiative resulted in a final report in 

May 2013 recommending that regional enforcement staff conduct 

periodic audits of nonescalated enforcement actions.

• “Minor” ROP realignment. According to NRC officials, in 2006, NRC 

undertook an internal review of substantive cross-cutting issues 

across regions. The review resulted, in part, from complaints from 

industry that some plants were assigned cross-cutting aspects more 

frequently than others, especially in Region IV. NRC officials stated 

that Region IV led the review, which found that, while each region was 

following guidance, it also had its own culture and exercised some 

 

36NRC, OIG-08-A-17. 
 

Page 40 GAO-13-743 Oversight of Reactor Safety

subjective judgment. Therefore, the regions had some differences in 

interpretations, especially in their approaches to determining what 

constitutes a more-than-minor performance deficiency. In response, 

according to NRC officials, NRC conducted additional training but 

made no key changes to its ROP procedures.

In addition, NRC officials said that several of the agency’s formal ongoing 

reviews examine, to a limited extent, the consistency of its oversight 

processes:

• ROP Enhancement Project. In January 2013, as part of its annual 

self-assessment, NRC initiated the ROP Enhancement Project, which 

is intended as a long-term, strategic review of how it implements the 

ROP, among other things. NRC officials said that they expect this 

review to consider regional differences in the number of nonescalated 

findings, but according to an NRC status report, the assessment 

portion of the review has been delayed pending the outcome of an 

independent review of the ROP. NRC officials told us they expect 

procedural changes in response to the project to begin in 2014.

• ROP realignment process. In 2009, NRC implemented the ROP 

realignment process, a biennial review of its ROP baseline inspection 

procedures and inspection hours per procedure to ensure the most 

effective overall applications of inspection resources. NRC officials 

told us that this effort also touches on the consistency of how NRC 

implements its oversight of safety.

• Self-assessments. In 2001, NRC initiated annual self-assessments, 

which are to assess whether the ROP met its program goals and 

achieved its intended outcomes. NRC officials said that these selfassessments consider, to a limited extent, issues related to 

consistency, and according to NRC guidance, these self-assessments 

assess, in part, the extent to which NRC implements the ROP 

predictably.

However, the biennial ROP realignment process evaluates specific 

inspection procedures agencywide to ensure the most effective 

application of inspection resources, and according to NRC guidance, the 

self-assessments are not intended to audit the performance of the regions 

in implementing the ROP. Thus even with taking these steps to examine 

the consistency of its oversight processes, agency officials told us that 

NRC has not conducted a comprehensive analysis of the causes of the 

differences in the number of nonescalated findings identified across 

regions. Under federal standards for internal control, managers at the 
 

Page 41 GAO-13-743 Oversight of Reactor Safety

functional or activity level are to compare actual performance to planned 

or expected results throughout the organization and analyze significant 

differences.37 Without conducting such an analysis, NRC (1) does not 

know whether its regional offices or individual inspectors are applying 

regulations and guidance consistently in handling findings; or (2) cannot 

identify other factors that may have led to the differences.

Moreover, NRC’s 2008 to 2013 strategic plan states that the agency will 

establish and maintain stable and predictable regulatory programs and 

policies, and other NRC guidance states that the agency should use an 

objective, understandable, and predictable process to ensure that 

licensees fulfill their responsibility for ensuring the safe operation of 

commercial nuclear reactors. In particular, the agency’s enforcement 

manual emphasizes that staff should ensure that oversight efforts are 

objective and consistent. NRC is aware of differences across regions in 

identifying and resolving the findings that result from its oversight 

processes. However, without a comprehensive analysis of the causes of 

the differences in the number of nonescalated findings across regions, 

NRC cannot ensure that oversight efforts are objective and consistent, as 

specified in its enforcement manual, or know whether its regional offices 

or individual inspectors are applying regulations and guidance 

consistently.

NRC maintains records of its identification of and response to both 

escalated and nonescalated findings; however, these records are difficult 

to search for several reasons, and limited our ability to assess the extent 

to which NRC followed its procedures and, thus, consistently identified 

and resolved findings. According to NRC’s strategic plan, the agency’s 

organizational values include openness in communication and decision 

making. Furthermore, NRC guidance states that the agency 

communicates ROP results to, in part, keep external stakeholders 

informed of licensee performance and to enhance confidence that NRC is 

accomplishing its mission. NRC’s enforcement policy states that its 

mission includes promoting the transparency and openness of its 

enforcement program for all stakeholders.

 

37GAO, Standards for Internal Control in the Federal Government, GAO/AIMD-00-21.3.1 

(Washington, D.C.: Nov.1999). 

Records of Findings Are 

Difficult to Search and 

Assess
 

Page 42 GAO-13-743 Oversight of Reactor Safety

In reviewing NRC’s findings data since the ROP’s inception in 2000 

through 2012, we found that NRC has recorded approximately 12,000 

findings––about 2 percent of which were escalated. For 95 percent of the 

findings data we reviewed—including numerous escalated findings—NRC 

did not include a case number for each finding, making it time-consuming 

and difficult to locate all documents associated with a single finding. 

According to NRC officials, although nonescalated findings each have a 

unique identifier, only the case number is searchable in NRC’s 

databases. NRC officials told us that, as a matter of policy, they do not 

assign case numbers to nonescalated findings; only to escalated findings. 

However, our review of NRC findings data also found 38 instances of 

escalated findings that had not been assigned a case number, during this 

period.

The information NRC makes publicly available on its identification of and 

response to findings was also difficult to use. NRC communicates the 

results of much of its oversight process to the public through its ROP 

website. This website makes available inspection reports, assessment 

letters, and other general materials related to NRC’s oversight process. 

NRC also provides a quarterly summary of every reactor’s performance, 

consisting of its inspection findings, the color of each performance 

indicator, and the reactors’ placement on the action matrix. However, to 

review this information, users must search a link for each plant on a 

quarterly basis. Since the ROP has been in place for more than 12 years, 

users may have to search dozens of links per plant and read many 

different documents to get complete information on individual findings. In 

addition, while descriptions in each document contain information about 

whether a finding was associated with a cross-cutting issue, the website 

itself does not provide information on those plants that had open 

substantive cross-cutting issues prior to the most recent assessment 

period. This information can be found on the website by linking to each 

plant’s individual assessment letters. NRC program officials 

acknowledged that it is difficult for users of the website to determine 

which plants had substantive cross-cutting issues open in the past. 

Therefore, it is difficult to identify, over time, those that NRC views as 

having potentially more significant problems.

NRC also makes publicly available its official recordkeeping system, the 

Agencywide Documents Access and Management System (ADAMS). 

ADAMS contains more than 730,000 documents dating back to 
 

Page 43 GAO-13-743 Oversight of Reactor Safety

November 1999,38 and several hundred new documents on NRC’s reactor 

oversight processes—as well as other NRC activities—are added each 

day. However, we found ADAMS difficult to navigate. For example, using 

an inspection report as our starting point, we tried to retrace the steps in 

NRC’s response to the findings in the report using ADAMS; however, our 

searches resulted in numerous unrelated documents as well as 

documents with ambiguous names that had to be opened to determine 

whether they were related to the findings. Further, we had no way of 

determining whether our searches returned a complete set of results and 

had identified all the documents related to the findings. NRC officials 

acknowledged that, while ADAMS can be used to conduct searches on 

specific words or other criteria, such searches may produce a large 

number of unrelated documents, in addition to relevant documents.

We asked NRC officials how we could track a finding through its 

resolution, and they told us that doing so is difficult because findings and

enforcement actions are recorded in separate databases and that these 

databases are not linked. The officials told us that users with expertise in 

both databases’ search capabilities could craft a data search that linked 

inspection and enforcement records. However, because external users 

cannot craft such a search, they cannot independently track NRC’s 

identification and resolution of findings. NRC’s Chairman reiterated the 

agency’s support for transparency in prepared remarks before the annual 

NRC Regulatory Information Conference in March 2013, stating that the 

agency must have a continued commitment to openness and 

transparency and communicate with interested parties in ways they 

understand, and that clear, consistent communication is a principle of 

good regulation. However, because external users cannot independently 

track findings and their resolution, NRC cannot ensure that the public, 

Congress, and others are kept informed of licensee performance, as 

specified in guidance, to enhance confidence that NRC is accomplishing 

its mission and following its oversight processes.

 

38According to NRC officials, ADAMS contains some documents dating from before 

November 1999. 
 

Page 44 GAO-13-743 Oversight of Reactor Safety

NRC has both formal and informal methods for developing lessons 

learned to improve its oversight. Formal methods include agencywide 

programs and annual assessments. Informal methods include reaching 

out to peer groups and technical experts throughout the agency and 

accessing various agency databases. Although NRC guidance directs 

inspectors to use information in agency databases on past experiences to 

plan and conduct future inspection activities, inspectors face challenges 

accessing this information, which may limit their ability to use it.

NRC has ongoing, formal methods through which it collects and analyzes 

information to develop lessons learned for improving reactor oversight,

including the following agencywide programs:

• Operating Experience Program, which captures information on past 

experience based on inspection findings, events, and documents from 

industry, international nuclear agencies, and others to inform agency 

activities, including reactor inspections.

• Generic Issues Program, which tracks the status and resolutions of 

certain issues involving public health and safety, among other things, 

that could affect multiple entities under NRC jurisdiction, such as 

multiple commercial nuclear plants.

• ROP Feedback Program, which enables staff to submit comments 

and suggestions for improving the ROP based on their experiences; a 

central coordinator then directs these comments and suggestions to 

appropriate internal stakeholders for their review.

• Industry Trends Program, which monitors trends in indicators of 

industry performance. NRC evaluates any adverse trends to 

determine whether regulatory action is appropriate, and reports 

statistically significant trends to Congress.

• Monthly ROP working group meetings, which, according to NRC 

officials, NRC hosts and include the Nuclear Energy Institute, other 

industry representatives, and members of the public to discuss 

operating experience, inspection and performance assessment topics, 

potential changes and questions regarding performance indicator 

guidance, and other topics.

NRC Uses Numerous 

Methods to Develop 

Lessons Learned for 

Improving Oversight, 

but Challenges 

Remain in Accessing 

This Information

NRC Has Formal Methods 

for Collecting and 

Analyzing Information to 

Develop Lessons Learned
 

Page 45 GAO-13-743 Oversight of Reactor Safety

Additionally, according to NRC documents, NRC identifies lessons 

learned during its annual self-assessments to determine whether the 

ROP met goals and during its biennial ROP realignment process. For 

example, the ROP realignment process assesses feedback from 

inspectors and other NRC officials submitted through the ROP Feedback 

Program.

NRC has also undertaken special initiatives, such as the Fukushima 

Near-Term Task Force, to develop lessons learned from the accident at 

the Fukushima Daiichi Nuclear Power Plant in Japan that it could apply in 

its oversight of U.S. commercial nuclear power reactors. NRC convened 

the task force in 2011 to review its processes and regulations and 

determine whether lessons learned from the accident could inform its 

oversight processes. The task force concluded that a sequence of events 

like the Fukushima accident is unlikely to occur in the United States and 

that the continued operation of the nation’s commercial nuclear reactors 

does not pose an imminent risk to public health and safety. In addition, in 

2011 the task force made 12 recommendations, such as for NRC to 

clarify its regulatory framework and to require licensees to reevaluate and 

upgrade seismic and flooding protection of reactors and related 

equipment, strengthen capabilities at all reactors to withstand loss of 

electrical power, and take other actions to better protect their plants for a 

low-probability high-impact event. NRC accepted these recommendations 

and is in the process of implementing them (see app. X).

To follow up on the task force’s recommendations, in 2011, NRC 

established the Japan Lessons Learned Project Directorate—a group of 

over 20 NRC staff focused on implementing the lessons the task force 

identified. As part of this effort, NRC issued the following three orders to 

licensees in March 2012:

• improve capability to reliably and remotely monitor a wider range of 

conditions in the pools that contain spent fuel;39

 

39According to NRC information, every reactor site has at least one spent fuel pool into 

which fuel is placed for storage when it is removed from the reactor. Spent nuclear fuel 

refers to the bundles of uranium pellets encased in metal rods that have been used to 

power a nuclear reactor. Nuclear fuel loses efficiency over time and becomes unable to 

keep a nuclear reaction going. Periodically, about one-third of the fuel assemblies in a 

reactor are replaced. The nuclear reaction is stopped before the spent fuel is removed. 

But spent fuel still produces a lot of radiation and heat that must be managed to protect 

workers, the environment, and the public. 
 

Page 46 GAO-13-743 Oversight of Reactor Safety

• install vent systems for certain types of reactors to help prevent an 

excess buildup of hydrogen gas in the reactor buildings;

• enhance plants’ capabilities to maintain or restore core cooling, 

containment, and spent fuel pool cooling in the case of external 

events, such as an earthquake, that exceeds plants’ design 

capabilities.

NRC has also undertaken special initiatives in response to other unique 

events. For example, following the terrorist attacks on September 11, 

2001, NRC issued orders to licensees to enhance security measures to 

protect their reactors against an increased terrorism threat. In addition, in 

2002, after extensive degradation was discovered in the reactor vessel 

head at the Davis-Besse Nuclear Power Station in Ohio,40 NRC 

established a lessons learned task force and implemented many of its 

recommendations, including inspection procedure changes, enhancing its 

training program, and enhancing aspects of program management of the 

ROP. In May 2004, we made several recommendations to ensure that 

NRC addresses weaknesses that contributed to the Davis-Besse incident, 

which NRC has since implemented.41

NRC officials told us they also have more informal methods of collecting 

and analyzing information to develop lessons learned for improving 

reactor oversight. These informal means include reaching out to peers by 

telephone or e-mail, reaching out to regional or headquarters technical 

experts by telephone or e-mail, and informal gatherings during peer group 

meetings. For example, several resident inspectors told us they typically 

rely on a small group of fellow resident inspectors—often within the same 

branch overseeing a small number of plants—that they call periodically 

for updates on issues about which they are known to have prior 

experience.

 

40In March 2002, the most serious safety issue confronting the nation’s commercial 

nuclear power industry since Three Mile Island in 1979 was identified at the Davis-Besse 

plant in Ohio. After NRC allowed Davis-Besse to delay shutting down to inspect its reactor 

vessel for cracked tubing, the plant found that leakage from these tubes had caused 

extensive corrosion on the vessel head—a vital barrier preventing a radioactive release. 

41GAO, Nuclear Regulation: NRC Needs to More Aggressively and Comprehensively 

Resolve Issues Related to the Davis-Besse Nuclear Power Plant’s Shutdown, 

GAO-04-415 (Washington, D.C.: May 17, 2004). 

NRC Also Has Informal 

Methods for Collecting and 

Analyzing Information to 

Develop Lessons Learned
 

Page 47 GAO-13-743 Oversight of Reactor Safety

Inspectors told us that this type of outreach is especially important 

because it allows them to tap into other inspectors’ different experiences 

performing inspections and unique engineering expertise, given that 

resident inspectors come from a variety of engineering backgrounds. 

They also told us that this form of outreach can have a large influence on 

an inspector’s perspective about potential findings. Resident inspectors 

stated that they hold many informal discussions about lessons learned—

including sharing findings and reviewing changes to guidance—during 

their semiannual conferences with peers from across their region and 

during periodic visits to plants in other regions. NRC uses e-mail to 

distribute updates agencywide to its guidance and other policy statements 

on how to respond to lessons learned from different plants; however, 

according to inspectors, individual inspectors are responsible for 

incorporating these lessons learned into their oversight activities. In 

addition, according to NRC officials, each region determines its policy for 

e-mailing lessons learned from across the region or lessons learned with 

generic safety implications.

Although NRC guidance directs inspectors to use information on past 

experiences to plan and conduct future inspection activities, inspectors 

reported facing challenges in accessing information on past experiences, 

which may limit their ability to benefit from it. Under NRC guidance, 

inspectors are to use information on past plant experiences to prepare 

for, conduct, and document inspection activities. According to several 

inspectors, this information can be important for helping them choose 

inspection samples. Some inspectors we interviewed also stated that 

complete access to lessons learned from past plant experience is 

important to help them understand how NRC identified and resolved 

similar findings at other plants. In addition, industry representatives have 

stated that information on past experiences is beneficial because the 

industry is mature and there is a robust body of information on past 

experiences to learn from.

However, several NRC officials told us that the Reactor Operating 

Experience (ROE) Gateway—the agency’s internal website for the 

Operating Experience Program—and its search tools are not userfriendly. For example, they noted that the website provides links to more 

than a dozen search tools that all contain different information but does 

not explain how to use them. The officials said that providing instructions 

on how to use these tools would be helpful. Moreover, they told us that 

conducting a thorough search on a specific topic may require multiple 

keyword searches because different reactor equipment manufacturers 

Challenges Accessing 

Information on Past 

Experiences May Limit 

Inspectors’ Ability to 

Benefit from It
 

Page 48 GAO-13-743 Oversight of Reactor Safety

use different terms to describe the same type of equipment. For example, 

one manufacturer of commercial nuclear reactors calls certain pumps 

Auxiliary Feedwater Pumps, while another manufacturer calls them 

Emergency Feedwater Pumps, although both serve the same function—

removing heat from the reactor core when the plant experiences a 

complete loss of electrical power. As a result, inspectors and other 

agency users—including officials in the Operating Experience Program 

and technical specialists in the regional offices—have to conduct 

searches on several possible keywords in each of the various search 

tools. NRC officials told us the various terms used to describe the same 

equipment are not officially documented so staff must learn about them 

through their own experiences. For example, if an NRC official 

encountered a problem at a plant with a particular type of pump, using 

ROE Gateway to search for past experiences with this pump would 

require the official to know that this piece of equipment had more than 

one name and to search on both names in each of the search tools. At 

least one of the search tools on ROE Gateway now uses drop-down lists 

to eliminate the need to conduct multiple keyword searches, but the 

remaining search tools do not have this capability.

Given the challenges in navigating ROE Gateway, several resident 

inspectors told us they prefer contacting other inspectors informally or 

requesting assistance from NRC headquarters staff when trying to obtain 

information on past experience. They also told us they have developed 

informal “workarounds” to facilitate searches, such as creating and 

sharing slide presentations with one another on keyword searches for 

various common inquiries. We have previously reported that informal 

relationships could end once personnel move to their next assignments.42

Thus an informal workaround cannot substitute for a user-friendly, 

searchable system. Without better search tools, inspectors could overly 

rely on information available through informal channels rather than the 

more comprehensive information located in agency databases. NRC 

guidance states that the efficient retrieval of information on past 

experiences is a basic requirement for meeting agencywide objectives 

and that such information is necessary to improve safety assessments 

and NRC decisions. Further, NRC officials noted that inspectors have 

varied backgrounds—that is, some may have expertise in mechanical 

 

42GAO, National Security: Key Challenges and Solutions to Strengthen Interagency 

Collaboration, GAO-10-822T (Washington, D.C.: June 9, 2010). 
 

Page 49 GAO-13-743 Oversight of Reactor Safety

engineering, while others have expertise in electrical engineering—

therefore they depend on NRC’s network of information to assess most 

issues and determine their importance. Past inspection results and 

related information are crucial parts of that network. NRC officials have 

acknowledged the difficulty of using ROE Gateway and noted at their 

most recent annual meeting in March 2013 that NRC has struggled to 

make information about past experiences accessible and distribute this 

information efficiently.

Several resident inspectors also told us that e-mail updates on past 

experiences could be managed more effectively. For example, once 

resident inspectors receive an e-mail, it is their responsibility to 

incorporate the information on past experiences at their plant. However, 

these resident inspectors noted that not all e-mail updates apply to all 

plants; for example, an update may apply to a certain type of reactor that 

is not in use at all plants. Some resident inspectors stated that one way to 

improve this system would be to prioritize the e-mail notices and indicate 

the extent to which they are relevant to the inspectors. In addition, some 

resident inspectors noted that nonurgent e-mail updates should be 

compiled, whereas now they are included alongside all e-mail updates.

Resident inspectors play an essential role in ensuring that reactors are 

operating safely because they have the opportunity and authority to, at 

any time, physically inspect plant equipment and operations, review plant 

records, and verify the accuracy of licensee-reported quantitative 

measures or indicators of plant performance. NRC officials told us that, 

because of their essential role in NRC’s oversight, inspectors need 

complete and efficient access to all the information necessary to do their 

jobs effectively. In addition, information on past plant performance and 

oversight actions helps inspectors know where to begin looking for 

potential issues of concern, and NRC guidance states that inspectors 

should consider this information in preparing for, conducting, and 

documenting inspection activities. For example, prior to the reactor 

damage found at the Davis-Besse plant in 2002, inspectors did not 

regularly focus on identifying reactor vessel head corrosion; however, as 

a result of this incident, NRC inspectors now focus on this issue regularly.

NRC oversees the safety of commercial nuclear power reactors using 

established, ongoing processes. NRC’s oversight has taken on even 

greater importance as many commercial reactors in the United States are 

operating under renewed licenses that will allow them to operate beyond 

their initial 40-year operating periods. NRC is aware of differences across 

Conclusions 
 

Page 50 GAO-13-743 Oversight of Reactor Safety

its regions in the identification and resolution of findings that result from 

its oversight processes and has taken some steps to address them, 

particularly in the differences of nonescalated findings. However, while 

NRC has undertaken these steps, NRC has not conducted a 

comprehensive analysis of these differences consistent with federal 

standards of internal control. Without a comprehensive analysis of the 

differences in the number of nonescalated findings across regions and 

their causes, NRC cannot ensure that oversight efforts are objective and 

consistent, as specified in its enforcement manual, or know whether its 

regional offices or individual inspectors are not applying regulations and 

guidance consistently.

Furthermore, NRC policy states that the agency’s mission includes 

promoting transparency and openness for all stakeholders. As part of 

such transparency, NRC makes publicly available its official 

recordkeeping system, which contains hundreds of thousands of 

documents on NRC’s reactor oversight processes—as well as other NRC 

activities. However, we found this system difficult to navigate, and 

external users cannot use it to independently track findings, all 

documents related to the findings, and the findings’ resolution. As a 

result, NRC cannot ensure that the public, Congress, and others are kept 

informed of licensee performance, as specified in agency guidance, to 

enhance public confidence that NRC is accomplishing its mission and 

following its oversight processes.

Information about reactor operators’ past experiences and NRC’s 

responses to them also plays a key role in NRC’s oversight processes. 

Although NRC guidance directs inspectors to use information on past 

experiences to plan and conduct future inspection activities, inspectors 

reported facing challenges in accessing information on past experiences, 

which may limit their ability to benefit from this information. Given the 

challenges in navigating ROE Gateway, several resident inspectors 

reported contacting other inspectors informally or requesting assistance 

from NRC headquarters staff when trying to obtain information on past 

experience. Without better search tools, inspectors could overly rely on 

information available through informal channels. NRC guidance states 

that the efficient retrieval of information on past experiences is a basic

requirement for meeting agencywide objectives and that such information 

is necessary to improve safety assessments and NRC decisions.
 

Page 51 GAO-13-743 Oversight of Reactor Safety

To improve NRC’s oversight processes, we are making three 

recommendations to the NRC Commissioners:

• To better meet its goal of implementing objective and consistent 

oversight, direct agency managers to conduct a comprehensive 

analysis of the causes of the differences in the identification and 

resolution of findings.

• To improve transparency and better enable the public, Congress, and 

others to independently track findings, all documents related to the 

findings, and the findings’ resolution, direct the agency to either

modify NRC’s publicly available recordkeeping system to do so or 

develop a publicly accessible tool that does so.

• To help NRC staff more efficiently use past experiences in their 

oversight activities, direct agency officials to evaluate the challenges 

inspectors face in retrieving all relevant information on plant 

performance and previous oversight activities and improve its systems 

accordingly to address these challenges.

We provided a draft copy of this report to the Executive Director for 

Operations of NRC for review and comment. In its written comments, 

reproduced in appendix XI, NRC generally agreed with our findings and 

recommendations. NRC also provided technical comments, which we 

incorporated, as appropriate.

To address our first recommendation, NRC stated that it will revisit its 

initiatives with respect to implementation of the ROP and the 

nonescalated enforcement process to identify potential enhancements. 

For our second recommendation, NRC stated that it will identify ways to 

improve transparency and track documents related to inspection findings 

through improved tools to facilitate public access to inspection 

information. To address our third recommendation, NRC stated that it will 

make plant performance and oversight information more readily 

searchable and available to inspection staff and other NRC personnel.

In addition, in its comment letter, concerning regional differences, NRC 

stated that, although it believes adequate internal controls to ensure 

alignment between regions exist through program office oversight and 

audits, among other things, it agrees to seek enhancements in this area. 

NRC raised concerns about repeated references in the draft report to the 

use of professional judgment by NRC staff members in the ROP and 

Traditional Enforcement Process, stating that these references imply a 

Recommendations for 

Executive Action

Agency Comments
 

Page 52 GAO-13-743 Oversight of Reactor Safety

high degree of subjectivity in the agency’s implementation of its oversight 

processes and that its use of professional judgment is excessive or 

inconsistent. NRC stated that it believes its use of professional judgment 

is limited and controlled through its guidance. As stated in the report, we 

agree that NRC’s oversight processes are largely prescribed in guidance.

As agreed with your offices, unless you publicly announce the contents of 

this report earlier, we plan no further distribution until 30 days from the 

report date. At that time, we will send copies to the appropriate 

congressional committees, NRC, and other interested parties. In addition, 

this report also will be available at no charge on the GAO website at 

http://www.gao.gov.

If you or your staff members have any questions about this report, please 

contact me at (202) 512-3841 or ruscof@gao.gov. Contact points for our

Offices of Congressional Relations and Public Affairs may be found on 

the last page of this report. Key contributors to this report are listed in 

appendix XII.

Frank Rusco

Director, Natural Resources and Environment
Appendix I: Objectives, Scope, and 

Methodology

Page 53 GAO-13-743 Oversight of Reactor Safety

Our objectives were to (1) describe how the Nuclear Regulatory 

Commission (NRC) implements its processes for overseeing the safety of 

commercial nuclear power reactors; (2) evaluate the extent to which NRC 

consistently identifies and resolves findings through these processes; and 

(3) describe NRC’s methods for developing lessons learned to improve its 

oversight and challenges, if any, NRC faces in doing so.

To describe how NRC implements its processes for overseeing the safety 

of commercial nuclear power reactors, we reviewed relevant federal laws, 

regulations, and guidance, including NRC’s Enforcement Policy, 

Enforcement Manual, and Inspection Manual, among many other 

guidance documents. We also reviewed previous GAO and NRC 

Inspector General reports that identified safety-related challenges or 

made recommendations for improving NRC’s oversight of reactor safety, 

and we reviewed documentation of NRC efforts to update its oversight 

processes. We interviewed NRC headquarters officials from the Offices of 

Nuclear Reactor Regulation, Enforcement, Nuclear Regulatory Research, 

and the Inspector General. We also conducted semistructured interviews 

with officials from all four NRC regions, as well as resident inspectors and 

plant representatives during site visits to five nuclear power plants 

(discussed further below). During these semistructured interviews, we 

asked NRC officials and plant representatives about (1) the 

implementation and consistency of NRC’s oversight activities, (2) the 

extent to which NRC’s oversight includes lessons learned from operating 

experience and safety-related findings, (3) the level of communication 

between NRC and plant representatives, (4) NRC actions taken to identify 

and resolve specific safety-related findings, and (5) site-specific 

considerations related to safety. In addition, we interviewed 

representatives from the Nuclear Energy Institute, the Union of 

Concerned Scientists, academic specialists, and other groups and 

individuals to discuss their views and gain additional insight into NRC’s 

oversight of reactor safety.

We made site visits to a nonprobability sample of five nuclear power 

plants. During these site visits, we held separate semistructured 

interviews with NRC resident inspectors and plant representatives, 

including licensee managers and plant operators, to obtain in-depth 

information about their experiences with NRC’s oversight process. We 

also toured several of the sites with the resident inspectors and observed 

areas and equipment related to previous findings, in addition to current 

issues of concern being assessed for potential findings. Because this was 

a nonprobability sample, the information we gathered from these site 

visits is not generalizable to all U.S. nuclear power plants but provides 

Appendix I: Objectives, Scope, and 

Methodology 
Appendix I: Objectives, Scope, and 

Methodology

Page 54 GAO-13-743 Oversight of Reactor Safety

important illustrative information for understanding the oversight of 

reactors at the selected plants. To select these sites, we applied the 

following criteria to capture a variety of characteristics: whether NRC’s 

assessment of the plant’s safety significance has changed in the past 2 

years, and whether, collectively, they were located in multiple NRC 

regions, operated by different licensees, and included different reactor 

types. In Region I, we visited the Calvert Cliffs Nuclear Power Plant in 

Maryland. In Region III, we visited the Byron Station in Illinois and 

Palisades Nuclear Plant in Michigan. In Region IV, we visited the Cooper 

Nuclear Station in Nebraska and Wolf Creek Generating Station in 

Kansas.

To evaluate the extent to which NRC consistently identifies and resolves 

findings through its processes for overseeing reactor safety, in addition to 

reviewing NRC guidance and interviewing NRC officials (as described 

above), we analyzed the nonsecurity findings and violations for all plants 

from the beginning of NRC’s Reactor Oversight Process in 2000 through 

2012.1 We obtained these data from NRC’s Reactor Program System 

database. We also analyzed separate NRC data on (1) the number of 

inspection hours from 2000 through 2012, (2) descriptive information 

about licensees, (3) substantive cross-cutting issues assigned to each 

plant from 2001 through 2012, and (4) the quarterly action matrix 

assessment for each reactor from 2001 through 2012. In order to assess 

the reliability of the data we analyzed, we reviewed database 

documentation, interviewed NRC officials familiar with the data, and 

conducted electronic tests of the data, looking for missing values, outliers, 

or other anomalies. We determined that the data were sufficiently reliable 

for our purposes. We also searched documents in NRC’s publicly 

available, web-based Agencywide Documents Access and Management 

System (ADAMS) database, which is the official recordkeeping system for 

all NRC documents.

To describe NRC’s methods for developing lessons learned to improve its 

oversight and challenges, if any, NRC faces in doing so, in addition to 

reviewing NRC guidance and interviewing NRC officials (as described 

above), we observed NRC headquarters officials demonstrating the 

various computer-based tools they use for communicating and providing 

 

1

We did not analyze security findings because they are outside the scope of this review, 

which focuses on NRC’s oversight of reactor safety, not security. 
Appendix I: Objectives, Scope, and 

Methodology

Page 55 GAO-13-743 Oversight of Reactor Safety

access to information on past experiences and lessons learned. We also 

observed resident inspectors, during our site visits to nuclear power 

plants, demonstrating how they access agency information on past 

experiences and lessons learned. In addition, we observed meetings 

where NRC staff updated the Commission and the public on the status of 

the development and implementation of lessons learned from the 

Fukushima disaster, among other safety-related concerns.

We conducted this performance audit from May 2012 to September 2013 

in accordance with generally accepted government auditing standards. 

Those standards require that we plan and perform the audit to obtain 

sufficient, appropriate evidence to provide a reasonable basis for our 

findings and conclusions based on our audit objectives. We believe that 

the evidence obtained provides a reasonable basis for our findings and 

conclusions based on our audit objectives.
Appendix II: Summary Information about 

Operating Nuclear Power Reactors in the 

United States

Page 56 GAO-13-743 Oversight of Reactor Safety

This appendix provides summary information about the 104 commercial 

nuclear power reactors at 65 power plants for which we analyzed NRC

data and that were operating in the United States through the end of 

calendar year 2012. For each reactor, table 14 lists the NRC region, 

state, date commercial operation began, and licensee.

Table 14: Summary Information about Operating Commercial Nuclear Power Reactors 

Reactor State 

Date commercial 

operation began Licensee 

NRC Region I

Beaver Valley Power Station, Unit 1 Pennsylvania Oct. 1976 First Energy Nuclear Operating Co.

Beaver Valley Power Station, Unit 2 Pennsylvania Nov.1987 First Energy Nuclear Operating Co.

Calvert Cliffs Nuclear Power Plant, Unit 1 Maryland May 1975 Calvert Cliffs Nuclear Power Plant Inc.

Calvert Cliffs Nuclear Power Plant, Unit 2 Maryland Apr. 1977 Calvert Cliffs Nuclear Power Plant Inc.

Hope Creek Generating Station, Unit 1 New Jersey Dec. 1986 PSEG Nuclear, LLC

Indian Point Nuclear Generating, Unit 2 New York Aug. 1974 Entergy Nuclear Operations, Inc.

Indian Point Nuclear Generating, Unit 3 New York Aug. 1976 Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant New York July 1975 Entergy Nuclear Operations, Inc.

Limerick Generating Station, Unit 1 Pennsylvania Feb. 1986 Exelon Generation Co., LLC

Limerick Generating Station, Unit 2 Pennsylvania Jan. 1990 Exelon Generation Co., LLC

Millstone Power Station, Unit 2 Connecticut Dec. 1975 Dominion Nuclear Connecticut, Inc.

Millstone Power Station, Unit 3 Connecticut Apr. 1986 Dominion Nuclear Connecticut, Inc.

Nine Mile Point Nuclear Station, Unit 1 New York Dec. 1969 Nine Mile Point Nuclear Station, LLC

Nine Mile Point Nuclear Station, Unit 2 New York Mar. 1988 Nine Mile Point Nuclear Station, LLC

Oyster Creek Nuclear Generating Station, Unit 1 New Jersey Dec. 1969 Exelon Generation Co., LLC

Peach Bottom Atomic Power Station, Unit 2 Pennsylvania July 1974 Exelon Generation Co., LLC

Peach Bottom Atomic Power Station, Unit 3 Pennsylvania Dec. 1974 Exelon Generation Co., LLC

Pilgrim Nuclear Power Station Massachusetts Dec .1972 Entergy Nuclear Operations, Inc.

R.E. Ginna Nuclear Power Plant New York July 1970 R.E. Ginna Nuclear Power Plant, LLC

Salem Nuclear Generating Station, Unit 1 New Jersey June 1977 PSEG Nuclear, LLC 

Salem Nuclear Generating Station, Unit 2 New Jersey Oct. 1981 PSEG Nuclear, LLC

Seabrook Station, Unit 1 New Hampshire Aug. 1990 FPL Energy Seabrook, LLC

Susquehanna Steam Electric Station, Unit 1 Pennsylvania June 1983 PPL Susquehanna, LLC

Susquehanna Steam Electric Station, Unit 2 Pennsylvania Feb. 1985 PPL Susquehanna, LLC

Three Mile Island Nuclear Station, Unit 1 Pennsylvania Sept. 1974 Exelon Generation Co., LLC

Vermont Yankee Nuclear Power Plant, Unit 1 Vermont Nov. 1972 Entergy Nuclear Operations, Inc.

Appendix II: Summary Information about 

Operating Nuclear Power Reactors in the 

United States
Appendix II: Summary Information about 

Operating Nuclear Power Reactors in the 

United States

Page 57 GAO-13-743 Oversight of Reactor Safety

Reactor State 

Date commercial 

operation began Licensee 

NRC Region II 

Browns Ferry Nuclear Plant, Unit 1 Alabama Aug. 1974 Tennessee Valley Authority 

Browns Ferry Nuclear Plant, Unit 2 Alabama Mar. 1975 Tennessee Valley Authority 

Browns Ferry Nuclear Plant, Unit 3 Alabama Mar. 1977 Tennessee Valley Authority 

Brunswick Steam Electric Plant, Unit 1 North Carolina Mar. 1977 Carolina Power & Light Co. 

Brunswick Steam Electric Plant, Unit 2 North Carolina Nov. 1975 Carolina Power & Light Co. 

Catawba Nuclear Station, Unit 1 South Carolina June 1985 Duke Energy Carolinas, LLC 

Catawba Nuclear Station, Unit 2 South Carolina Aug. 1986 Duke Energy Carolinas, LLC 

Crystal River Nuclear Generating Plant, Unit 3 Florida Mar. 1977 Florida Power Corp. 

Edwin I. Hatch Nuclear Plant, Unit 1 Georgia Dec. 1975 Southern Nuclear Operating Co. 

Edwin I. Hatch Nuclear Plant, Unit 2 Georgia Sept. 1979 Southern Nuclear Operating Co. 

H. B. Robinson Steam Electric Plant, Unit 2 South Carolina Mar. 1971 Carolina Power & Light Co. 

Joseph M. Farley Nuclear Plant, Unit 1 Alabama Dec. 1977 Southern Nuclear Operating Co. 

Joseph M. Farley Nuclear Plant, Unit 2 Alabama July 1981 Southern Nuclear Operating Co. 

McGuire Nuclear Station, Unit 1 North Carolina Dec. 1981 Duke Energy Carolinas, LLC

McGuire Nuclear Station, Unit 2 North Carolina Mar. 1984 Duke Energy Carolinas, LLC

North Anna Power Station, Unit 1 Virginia June 1978 Virginia Electric & Power Co. 

North Anna Power Station, Unit 2 Virginia Dec. 1980 Virginia Electric & Power Co. 

Oconee Nuclear Station, Unit 1 South Carolina July 1973 Duke Energy Carolinas, LLC

Oconee Nuclear Station, Unit 2 South Carolina Sept. 1974 Duke Energy Carolinas, LLC

Oconee Nuclear Station, Unit 3 South Carolina Dec. 1974 Duke Energy Carolinas, LLC

Sequoyah Nuclear Plant, Unit 1 Tennessee July 1981 Tennessee Valley Authority 

Sequoyah Nuclear Plant, Unit 2 Tennessee June 1982 Tennessee Valley Authority 

Shearon Harris Nuclear Power Plant, Unit 1 North Carolina May 1987 Carolina Power & Light Co. 

St. Lucie Plant, Unit 1 Florida Dec. 1976 Florida Power & Light Co. 

St. Lucie Plant, Unit 2 Florida Aug. 1983 Florida Power & Light Co. 

Surry Nuclear Power Station, Unit 1 Virginia Dec. 1972 Virginia Electric & Power Co. 

Surry Nuclear Power Station, Unit 2 Virginia May 1973 Virginia Electric & Power Co. 

Turkey Point Nuclear Generating, Unit 3 Florida Dec. 1972 Florida Power & Light Co. 

Turkey Point Nuclear Generating, Unit 4 Florida Sept. 1973 Florida Power & Light Co. 

Virgil C. Summer Nuclear Station, Unit 1 South Carolina Jan. 1984 South Carolina Electric & Gas Co. 

Vogtle Electric Generating Plant, Unit 1 Georgia June 1987 Southern Nuclear Operating Co. 

Vogtle Electric Generating Plant, Unit 2 Georgia May 1989 Southern Nuclear Operating Co. 

Watts Bar Nuclear Plant, Unit 1 Tennessee May 1996 Tennessee Valley Authority 
Appendix II: Summary Information about 

Operating Nuclear Power Reactors in the 

United States

Page 58 GAO-13-743 Oversight of Reactor Safety

Reactor State 

Date commercial 

operation began Licensee 

NRC Region III 

Braidwood Station, Unit 1 Illinois July 1988 Exelon Generation Co., LLC 

Braidwood Station, Unit 2 Illinois Oct. 1988 Exelon Generation Co., LLC 

Byron Station, Unit 1 Illinois Sept. 1985 Exelon Generation Co., LLC 

Byron Station, Unit 2 Illinois Aug. 1987 Exelon Generation Co., LLC 

Clinton Power Station, Unit 1 Illinois Nov. 1987 Exelon Generation Co., LLC 

Davis-Besse Nuclear Power Station, Unit 1 Ohio July 1978 First Energy Nuclear Operating Co. 

Donald C. Cook Nuclear Power Plant, Unit 1 Michigan Aug. 1975 Indiana Michigan Power Co. 

Donald C. Cook Nuclear Power Plant, Unit 2 Michigan July 1978 Indiana Michigan Power Co. 

Dresden Nuclear Power Station, Unit 2 Illinois June 1970 Exelon Generation Co., LLC 

Dresden Nuclear Power Station, Unit 3 Illinois Nov. 1971 Exelon Generation Co., LLC 

Duane Arnold Energy Center Iowa Feb. 1975 FPL Energy Duane Arnold, LLC 

Fermi, Unit 2 Michigan Jan. 1988 The Detroit Edison Co. 

Kewaunee Power Station Wisconsin June 1974 Dominion Energy Kewaunee, Inc. 

LaSalle County Station, Unit 1 Illinois Jan. 1984 Exelon Generation Co., LLC 

LaSalle County Station, Unit 2 Illinois Oct. 1984 Exelon Generation Co., LLC 

Monticello Nuclear Generating Plant, Unit 1 Minnesota June 1971 Northern States Power Company 

Palisades Nuclear Plant Michigan Dec. 1971 Entergy Nuclear Operations, Inc. 

Perry Nuclear Power Plant, Unit 1 Ohio Nov. 1987 First Energy Nuclear Operating Co. 

Point Beach Nuclear Plant, Unit 1 Wisconsin Dec. 1970 FPL Energy Duane Arnold, LLC 

Point Beach Nuclear Plant, Unit 2 Wisconsin Oct. 1972 FPL Energy Duane Arnold, LLC 

Prairie Island Nuclear Generating Plant, Unit 1 Minnesota Dec. 1973 Northern States Power Co. Minnesota 

Prairie Island Nuclear Generating Plant, Unit 2 Minnesota Dec. 1974 Northern States Power Co. Minnesota 

Quad Cities Nuclear Power Station, Unit 1 Illinois Feb. 1973 Exelon Generation Co., LLC 

Quad Cities Nuclear Power Station, Unit 2 Illinois Mar. 1973 Exelon Generation Co., LLC 

NRC Region IV

Arkansas Nuclear One, Unit 1 Arkansas Dec. 1974 Entergy Nuclear Operations, Inc. 

Arkansas Nuclear One, Unit 2 Arkansas Mar. 1980 Entergy Nuclear Operations, Inc. 

Callaway Plant Missouri Dec. 1984 Union Electric Co. 

Columbia Generating Station, Unit 2 Washington Dec. 1984 Energy Northwest 

Comanche Peak Steam Electric Station, Unit 1 Texas Aug. 1990 Luminant Generation Co., LLC 

Comanche Peak Steam Electric Station, Unit 2 Texas Aug. 1993 Luminant Generation Co., LLC 

Cooper Nuclear Station Nebraska July 1974 Nebraska Public Power District 

Diablo Canyon Nuclear Power Plant, Unit 1 California May 1985 Pacific Gas & Electric Co. 

Diablo Canyon Nuclear Power Plant, Unit 2 California Mar. 1986 Pacific Gas & Electric Co. 

Fort Calhoun Station, Unit 1 Nebraska Sept. 1973 Omaha Public Power District 
Appendix II: Summary Information about 

Operating Nuclear Power Reactors in the 

United States

Page 59 GAO-13-743 Oversight of Reactor Safety

Reactor State 

Date commercial 

operation began Licensee 

Grand Gulf Nuclear Station, Unit 1 Mississippi July 1985 Entergy Nuclear Operations, Inc. 

Palo Verde Nuclear Generating Station, Unit 1 Arizona Jan. 1986 Arizona Public Service Company 

Palo Verde Nuclear Generating Station, Unit 2 Arizona Sept. 1986 Arizona Public Service Company 

Palo Verde Nuclear Generating Station, Unit 3 Arizona Jan. 1988 Arizona Public Service Company 

River Bend Station, Unit 1 Louisiana June 1986 Entergy Nuclear Operations, Inc. 

San Onofre Nuclear Generating Station, Unit 2 California Aug. 1983 Southern California Edison Co. 

San Onofre Nuclear Generating Station, Unit 3 California Apr. 1984 Southern California Edison Co. 

South Texas Project, Unit 1 Texas Aug. 1988 STP Nuclear Operating Co. 

South Texas Project, Unit 2 Texas June 1989 STP Nuclear Operating Co. 

Waterford Steam Electric Station, Unit 3 Louisiana Sept. 1985 Entergy Nuclear Operations, Inc. 

Wolf Creek Generating Station, Unit 1 Kansas Sept. 1985 Wolf Creek Nuclear Operating Corp. 

Source: NRC.

Note: This table includes data on the 104 commercial nuclear power reactors operating in the United 

States through the end of calendar year 2012. In February 2013, the owner of the Crystal River 

Nuclear Plant in Florida permanently shut down that site’s reactor; in May 2013, the owner of the 

Kewaunee Power Station in Wisconsin permanently shut down that site’s reactor; and in June 2013, 

the owner of the San Onofre Nuclear Generating Station in California permanently shut down that 

site’s two reactors. These actions reduced the number of operating commercial nuclear power 

reactors in the United States from 104 to 100.
Appendix III: NRC Safety Cornerstones and 

Related Performance Indicators

Page 60 GAO-13-743 Oversight of Reactor Safety

Initiating Events cornerstone – This cornerstone focuses on operations and events at a nuclear plant that could lead to a possible 

accident, if plant safety systems did not intervene. These events could include equipment failures leading to a plant shutdown, 

shutdowns with unexpected complications, or large changes in the plant’s power output. (In this table, “scram” refers to automatic and 

manual reactor shutdowns.) 

Performance indicator Objective 

Unplanned scrams per 7,000 critical hours This performance indicator monitors the number of unplanned scrams. It measures the 

rate of scrams per year of operation at power and provides an indication of initiating 

event frequency. The value of 7,000 hours is used because it represents 1 year of 

reactor operation at 80 percent capacity. 

Unplanned scrams with complications This performance indicator monitors the subset of unplanned scrams while critical that 

require additional operator actions beyond that of the “normal” scram. Such events or 

conditions have the potential to present additional challenges to the plant operations 

staff and therefore, may be more risk-significant than uncomplicated scrams. 

Unplanned power changes per 7,000 

critical hours

This performance indicator monitors the number of unplanned power changes (excluding 

scrams) that could have, under other plant conditions, challenged safety functions. It 

may provide leading indication of risk-significant events but is not itself risk-significant. 

The indicator measures the number of plant power changes for a typical year of 

operation at power. 

Mitigating Systems cornerstone – This cornerstone measures the function of safety systems designed to prevent an accident or 

reduce the consequences of a possible accident. The equipment is checked by periodic testing and through actual performance.

Performance indicator Objective 

Mitigating system performance index – 

emergency alternating current power 

systems 

This performance indicator monitors the readiness of the emergency alternating current 

power systems to perform their safety functions in response to off-normal events or 

accidents. 

Mitigating system performance index – high 

pressure injection systems 

This performance indicator monitors the readiness of the high pressure injection systems 

to perform their safety functions in response to off-normal events or accidents. 

Mitigating system performance index – heat 

removal systems 

This performance indicator monitors the readiness of the heat removal systems to 

perform their safety functions in response to off-normal events or accidents. 

Mitigating system performance index – 

residual heat removal systems 

This performance indicator monitors the readiness of the residual heat removal systems 

to perform their safety functions in response to off-normal events or accidents. 

Mitigating system performance index – 

cooling water systems 

This performance indicator monitors the readiness of the cooling water support systems 

to perform their safety functions in response to off-normal events or accidents. 

Safety system functional failures This performance indicator monitors events or conditions that alone prevented, or could 

have prevented, the fulfillment of the safety function of structures or systems that are 

needed to: shut down the reactor and maintain it in a safe shutdown condition; remove 

residual heat; control the release of radioactive material; or mitigate the consequences of 

an accident. 

Appendix III: NRC Safety Cornerstones and 

Related Performance Indicators
Appendix III: NRC Safety Cornerstones and 

Related Performance Indicators

Page 61 GAO-13-743 Oversight of Reactor Safety

Barrier Integrity cornerstone – This cornerstone focuses on the physical barriers between the highly radioactive materials in fuel 

within the reactor and the public and the environment outside the plant. These barriers are the sealed rods containing the fuel pellets, 

the heavy steel reactor vessel and associated piping, and the reinforced concrete containment building surrounding the reactor. The 

integrity of the fuel rods, the vessel, and the piping is continuously checked for leakage, while the ability of the containment to prevent 

leakage is measured on a regular basis.

Performance indicator Objective 

Reactor coolant system specific activity This performance indicator monitors the integrity of the fuel cladding, the first of the three 

barriers to prevent the release of fission products. It measures the radioactivity in the 

reactor coolant system as an indication of functionality of the cladding. 

Reactor coolant system leakage This performance indicator monitors the integrity of the reactor coolant system pressure 

boundary, the second of the three barriers to prevent the release of fission products. It 

measures reactor coolant system identified leakage as a percentage of the technical 

specification allowable identified leakage to provide an indication of reactor coolant 

system integrity. 

Emergency Preparedness cornerstone –This cornerstone measures the effectiveness of plant staff in carrying out emergency plans 

for responding to a possible accident. Such emergency plans are tested every 2 years during emergency exercises involving the plant 

staff and local, state, and, in some cases, federal agencies.

Performance indicator Objective 

Drill/exercise performance This performance indicator monitors timely and accurate licensee performance in drills 

and exercises when presented with opportunities for classification of emergencies, 

notification of off-site authorities, and development of protective action 

recommendations. It is the ratio, in percent, of timely and accurate performance of those 

actions to total opportunities. 

Emergency response organization drill 

participation

This performance indicator tracks the participation of emergency response organization 

members assigned to fill key positions in performance enhancing experiences, and 

through linkage to the drill/exercise performance indicator ensures that the risksignificant aspects of classification, notification, and protective action recommendations 

development are evaluated and included in the performance indicator process. This 

indicator measures the percentage of emergency response organization members 

assigned to fill key positions who have participated recently in performance-enhancing 

experiences such as drills, exercises, or in an actual event. 

Alert and notification system reliability This performance indicator monitors the reliability of the off-site alert and notification 

system, a critical link for alerting and notifying the public of the need to take protective 

actions. It provides the percentage of the sirens that are capable of performing their 

safety function based on regularly scheduled tests. 

Occupational Radiation Safety cornerstone – NRC regulations set a limit on radiation doses received by plant workers, and this 

cornerstone monitors the effectiveness of the plant’s program to control and minimize those doses.

Performance indicator Objective 

Occupational exposure control 

effectiveness

This performance indicator monitors the control of access to and work activities within 

radiologically-significant areas of the plant and occurrences involving degradation or 

failure of radiation safety barriers that result in readily-identifiable unintended dose. 

Public Radiation Safety cornerstone – This cornerstone measures the procedures and systems designed to minimize radioactive 

releases from a nuclear plant during normal operations and to keep those releases within federal limits.

Performance indicator Objective 

“Radiological effluent technical 

specifications” / “offsite dose calculation 

manual” radiological effluent occurrence 

This performance indicator assesses the performance of the program to control 

radiological effluent occurrences, which are radiological releases that exceed pre-set 

limits for liquid and gaseous releases. 
Appendix III: NRC Safety Cornerstones and 

Related Performance Indicators

Page 62 GAO-13-743 Oversight of Reactor Safety

Physical Protection cornerstonea

Sources: GAO analysis of NRC and Nuclear Energy Institute information.

 – Nuclear plants are required to have well-trained security personnel and a variety of protective 

systems to guard vital plant equipment, as well as programs to assure that employees are constantly fit for duty through drug and 

alcohol testing. This cornerstone measures the effectiveness of the security and fitness-for-duty programs.

a

The Physical Protection cornerstone consists of physical security issues. Our review focuses on 

NRC’s oversight of the safe operation of reactors; therefore the physical security of plants and related 

performance indicators are outside the scope of this review.
Appendix IV: NRC Safety Cornerstones and 

Related Inspection Procedures

Page 63 GAO-13-743 Oversight of Reactor Safety

This appendix provides summary information about NRC’s inspection 

procedures. Table 15 lists the safety cornerstone, title, and objective for 

each of the 36 safety-related (nonsecurity) baseline inspection 

procedures. Table 16 lists the title and objective for each of the 3 

supplemental inspection procedures, and table 17 lists the title and 

objective for each of the 3 reactive inspection procedures.

Table 15: NRC Safety Cornerstones and Related Baseline Inspection Procedures 

Initiating Events cornerstone – This cornerstone focuses on operations and events at a nuclear plant that could lead to a possible 

accident, if plant safety systems did not intervene. These events could include equipment failures leading to a plant shutdown, 

shutdowns with unexpected complications, or large changes in the plant’s power output.

Mitigating Systems cornerstone – This cornerstone measures the function of safety systems designed to prevent an accident or 

reduce the consequences of a possible accident. The equipment is checked by periodic testing and through actual performance.

Barrier Integrity cornerstone – This cornerstone focuses on the physical barriers between the highly radioactive materials in fuel 

within the reactor and the public and the environment outside the plant. These barriers are the sealed rods containing the fuel pellets, 

the heavy steel reactor vessel and associated piping, and the reinforced concrete containment building surrounding the reactor. The 

integrity of the fuel rods, the vessel, and the piping is continuously checked for leakage, while the ability of the containment to prevent 

leakage is measured on a regular basis.

Inspection 

procedure Title Objective 

71111.01 Adverse weather 

protection

This inspection focuses on verifying that the design features and implementation of 

the licensee’s procedures protect mitigating systems from adverse weather effects. 

Adverse weather includes events such as high winds, hurricanes, electrical storms, 

tornadoes, and extreme high or low temperatures.

71111.04 Equipment alignment This inspection focuses on verifying equipment alignment, identifying any 

discrepancies that impact the function(s) of the system, and verifying that the 

licensee has properly identified and resolved equipment alignment problems. 

71111.05AQ Fire Protection 

(Annual/Quarterly)

This inspection focuses on evaluating the licensee’s fire protection program, including 

assessing the performance of the fire brigade and verifying the adequacy of its fire 

detection and suppression capability and controls for combustibles and ignition 

sources within the plant, among other things. 

71111.05T Fire Protection 

(Triennial)a This inspection focuses on evaluating the design, operational status, and material condition of the licensee’s fire protection program. In addition, inspection teams verify 

that certain mitigating strategies are feasible in light of operator training, maintenance 

of necessary equipment, and any plant modifications. This inspection is designed to 

complement Inspection Procedure 71111.05AQ.

71111.05XT Fire Protection – NFPA 

805 (Triennial)b This inspection focuses on evaluating the design, installation, operational status, testing, and material condition of the licensee’s fire protection program for plants, as 

well as the plant’s capability to meet requirements of the NRC-approved fire 

protection program, and goals, objectives, and criteria for fire safety of National Fire 

Protection Association (NFPA) Standard 805. In addition, inspection teams verify that 

certain mitigating strategies are feasible in light of operator training, maintenance of 

necessary equipment, and any plant modifications. This inspection is designed to 

complement Inspection Procedure 71111.05AQ.

 

71111.06 Flood Protection 

Measures

This inspection focuses on verifying that the licensee’s flooding mitigation plans and 

equipment are consistent with the licensee’s design requirements and the risk 

analysis assumptions. 

Appendix IV: NRC Safety Cornerstones and 

Related Inspection Procedures
Appendix IV: NRC Safety Cornerstones and 

Related Inspection Procedures

Page 64 GAO-13-743 Oversight of Reactor Safety

71111.07 Heat Sink Performance This inspection focuses on verifying that (1) potential heat exchanger deficiencies 

which could mask degraded performance are identified; (2) potential heat sink 

performance problems that have the potential to increase risk are identified; and (3) 

the licensee has adequately resolved them. 

71111.08 Inservice Inspection 

Activities

This inspection focuses on assessing the effectiveness of the licensee’s program for 

monitoring degradation of reactor coolant system boundary, risk-significant piping 

system boundaries, and the containment boundary. 

71111.11 Licensed Operator 

Requalification Program 

and Licensed Operator 

Performance 

This inspection focuses on evaluating licensed operator performance during facilityadministered requalification examinations, other examinations, facility training 

exercises, and during selected evolutions conducted in the actual plant/main control 

room. This inspection also reviews the licensee’s ability to identify and correct 

problems associated with licensed operator performance. 

71111.12 Maintenance 

Effectiveness

This inspection focuses on supplementing performance indicators by providing for 

independent oversight of licensee maintenance effectiveness. 

71111.13 Maintenance Risk 

Assessments and 

Emergent Work Control

This inspection focuses on verifying the performance and adequacy of risk 

assessments for planned or emergent maintenance activities, among other activities.

71111.15 Operability 

Determinations and 

Functionality 

Assessments

This inspection focuses on reviewing the operability or functionality assessments for 

certain components or systems to ensure that they remain capable of performing 

their design functions.

71111.17T Evaluations of Changes, 

Tests and Experiments 

and Permanent Plant 

Modifications

This inspection focuses on verifying that modifications to the plant have been 

adequately implemented, and that procedures and other documentation affected by 

the modifications have been adequately updated.

71111.18 Plant Modifications This inspection focuses on verifying that modifications to the plant have not affected 

the safety functions of important safety systems, have not degraded certain 

structures, systems, and components, and have not placed the plant in an unsafe 

condition. 

71111.19 Post-Maintenance 

Testing

This inspection focuses on verifying that post-maintenance test procedures and test 

activities are adequate to verify system operability, and functional capability. 

71111.20 Refueling and Other 

Outage Activities

This inspection focuses on evaluating licensee outage activities, including verifying 

that the licensee considered risk in developing outage schedules and adhered to 

certain risk-reduction methodologies. 

71111.21 Component Design 

Bases Inspection

This inspection focuses on obtaining reasonable assurance that risk-significant 

structures, systems, and components can adequately perform their design basis 

function. 

71111.22 Surveillance Testing This inspection focuses on verifying that surveillance testing of risk-significant 

structures, systems, and components are capable of performing their intended safety 

functions and assessing their operational readiness.
Appendix IV: NRC Safety Cornerstones and 

Related Inspection Procedures

Page 65 GAO-13-743 Oversight of Reactor Safety

Emergency Preparedness cornerstone –This cornerstone measures the effectiveness of plant staff in carrying out emergency plans 

for responding to a possible accident. Such emergency plans are tested every 2 years during emergency exercises involving the plant 

staff and local, state, and, in some cases, federal agencies.

Inspection

procedure Title Description 

71114.01 Exercise Evaluation This inspection focuses on evaluating the adequacy of the licensee’s conduct of each 

plant’s biennial exercise, which each plant is required to complete to demonstrate 

their capability to adequately perform key skills in certain functional areas to protect 

public health and safety in the event of a radiological emergency. 

71114.02 Alert and Notification 

System Testing

This inspection focuses on evaluating the licensee’s compliance with the testing and 

maintenance requirements for alert and notification systems. 

71114.03 Emergency Response 

Organization Staffing and 

Augmentation System

This inspection focuses on evaluating whether the licensee’s on-shift and 

augmentation staffing levels meet emergency response commitments.

71114.04 Emergency Action Level 

and Emergency Plan 

Changes

This inspection focuses on assessing whether any changes to the licensee’s 

emergency action level scheme have decreased the effectiveness of its emergency 

plan, and verifying that other emergency plan changes did not decrease its 

effectiveness. 

71114.05 Maintenance of 

Emergency 

Preparedness 

This inspection focuses on evaluating the efficacy of licensee efforts to maintain their 

emergency preparedness programs. 

71114.06 Drill Evaluation This inspection focuses on evaluating the licensee’s assessment of performance to 

identify weaknesses in emergency preparedness in selected drills and training 

evolutions and use of its corrective action program to correct these weaknesses. 

71114.07 Exercise Evaluation – 

Hostile Action Event

This inspection focuses on evaluating the adequacy of the licensee’s ability to 

implement mitigative measures in response to a simulated attack at the plant, 

coordinate required actions to successfully respond to and mitigate plant damage, as 

well as its capability to assess performance in order to identify and correct 

weaknesses. 

Occupational Radiation Safety cornerstone – NRC regulations set a limit on radiation doses received by plant workers, and this 

cornerstone monitors the effectiveness of the plant’s program to control and minimize those doses.

Public Radiation Safety cornerstone – This cornerstone measures the procedures and systems designed to minimize radioactive 

releases from a nuclear plant during normal operations and to keep those releases within federal limits.

Inspection

procedure Title Description 

71124.01 Radiological Hazard 

Assessment and 

Exposure Controls

This inspection focuses on assessing licensee performance in assessing the 

radiological hazards in the workplace and in implementing of appropriate radiation 

monitoring and exposure control measures. In addition, this inspection focuses on 

verifying that the licensee is properly identifying and reporting performance indicators 

for the Occupational Radiation Safety cornerstone. 

71124.02 Occupational ALARA 

Planning and Controls

This inspection focuses on assessing licensee performance in maintaining radiation 

exposures as low as reasonably achievable (ALARA). This inspection will determine 

whether the licensee’s ALARA program is effective. 

71124.03 In-Plant Airborne 

Radioactivity Control and 

Mitigation

This inspection focuses on verifying that in-plant airborne concentrations of 

radioactivity are as low as reasonably achievable to the extent necessary to validate 

plant operations and to verify that the practices and use of respiratory protection 

devices on site do not pose an undue risk to the wearer. 
Appendix IV: NRC Safety Cornerstones and 

Related Inspection Procedures

Page 66 GAO-13-743 Oversight of Reactor Safety

71124.04 Occupational Dose 

Assessment

This inspection focuses on determining the accuracy and operability of a licensee’s 

personal radiation monitoring equipment and the accuracy and effectiveness of its 

methods for calculating and monitoring radiological exposure to workers. 

71124.05 Radiation Monitoring 

Instrumentation

This inspection focuses on verifying that the licensee is ensuring the accuracy and 

operability of radiation monitoring instruments that are used to monitor areas, 

materials, and workers to ensure a radiologically safe work environment, among 

other things.

71124.06 Radioactive Gaseous 

and Liquid Effluent 

Treatment

This inspection focuses on ensuring that gaseous and liquid effluent processing 

systems are maintained so that radiological discharges are properly mitigated, 

monitored, and evaluated with regard to public exposure. 

71124.07 Radiological 

Environmental Monitoring 

Program 

This inspection focuses on verifying that the licensee quantifies the impact of 

radioactive effluent releases to the environment, that it sufficiently validates the 

integrity of its program that monitors radioactive gaseous and liquid effluent release, 

and that it implements this program consistently. 

71124.08 Radioactive Solid Waste 

Processing and 

Radioactive Material 

Handling, Storage, and 

Transportation

This inspection focuses on verifying the effectiveness of the licensee’s programs for 

processing, handling, storage, and transportation of radioactive material. 

Physical Protection cornerstonec

Other baseline procedures

 – Nuclear plants are required to have well-trained security personnel and a variety of protective 

systems to guard vital plant equipment, as well as programs to assure that employees are constantly fit for duty through drug and 

alcohol testing. This cornerstone measures the effectiveness of the security and fitness-for-duty programs.

Inspection

procedure Title Description 

71151 Performance Indicator 

Verification

This inspection focuses on performing a periodic review of performance indicator 

data to determine their accuracy and completeness. 

71152 Problem Identification 

and Resolution

This inspection focuses on (1) providing an early warning of potential performance 

issues that could result in crossing thresholds in the action matrix of NRC’s Reactor 

Oversight Process, (2) helping gauge NRC’s response when these thresholds are 

crossed, and (3) providing insights into whether licensees have established a safetyconscious work environment, among other things. 

71153 Follow-up of Events and 

Notices of Enforcement 

Discretion

This inspection focuses on evaluating licensee events and degraded conditions for 

plant status and mitigating actions to help determine the need for additional 

inspections. 

Source: GAO analysis of NRC information.

a

Traditionally, commercial nuclear power plants have followed a deterministic fire safety approach 

where plant operators are required to ensure that at least one system of electric cables and 

equipment is available to safely shut down a reactor if a fire occurs.

b

In 2004 NRC issued a regulation permitting plants to voluntarily transition from a deterministic fire 

safety approach to risk-informed fire protection requirements. Under this approach, a licensee adopts 

performance goals, objectives, and criteria for fire safety that are defined by the fire protection 

standard—NFPA-805—issued by the National Fire Protection Association (NFPA), an international 

nonprofit organization that develops, publishes, and disseminates fire prevention and safety 

standards.

c

This review focuses on NRC’s oversight of the safe operation of reactors; therefore the physical 

security of plants (Physical Protection cornerstone) is outside the scope of this review.
Appendix IV: NRC Safety Cornerstones and 

Related Inspection Procedures

Page 67 GAO-13-743 Oversight of Reactor Safety

Table 16: NRC Supplemental Inspection Procedures 

Inspection

procedure Title Description 

95001 Supplemental Inspection for One 

or Two White Inputs in a 

Strategic Performance Area

This inspection focuses on (1) providing assurance that the root causes and 

contributing causes of risk-significant performance issues are understood, (2) that 

the extent of condition and extent of cause of risk-significant performance issues are 

identified, and (3) that the licensee’s corrective actions for risk-significant 

performance issues are sufficient to address them. 

95002 Supplemental Inspection for One 

Degraded Cornerstone or any 

Three White Inputs in a Strategic 

Performance Area

This inspection focuses on (1) providing assurance that the cause(s) of risksignificant performance issues are understood, (2) independently assessing that the 

extent of the condition and extent of cause of risk-significant performance issues are 

identified, (3) independently determining if safety culture components caused or 

contributed to these performance issues, and (4) providing assurance that the 

licensee’s corrective actions are sufficient to prevent recurrence. 

95003 Supplemental Inspection for 

Repetitive Degraded 

Cornerstones, Multiple Degraded 

Cornerstones, Multiple Yellow 

Inputs or One Red Input

This inspection focuses on providing the NRC additional information to be used in

deciding whether the continued operation of the plant is acceptable and whether 

additional regulatory actions are necessary to stop the decline of a plant’s 

performance. In addition, it focuses on providing an independent assessment of the 

extent of risk-significant issues to determine whether an unacceptable margin of 

safety or security exists, among other things. 

Source: GAO analysis of NRC information.

Table 17: NRC Reactive Inspection Procedures 

Inspection

procedure Title Description 

93812 Special Inspection This inspection focuses on assessing a significant operational event and its causes 

with a special inspection team, including collecting, analyzing, and documenting 

information and evidence about the event, and evaluating the adequacy of the 

licensee’s response to the event.

The team conducting the inspection is composed of technical experts from the 

region in which the event took place. The special inspection team reports directly to 

the appropriate regional administrator. 

93800 Augmented Inspection Team This inspection focuses on assessing a significant operational event and its causes 

with an augmented inspection team, including collecting, analyzing, and 

documenting information and evidence about the event, and evaluating the 

adequacy of the licensee’s response to the event.

It is similar to an inspection performed by a special inspection team except that the 

team conducting an augmented inspection consists of technical experts from the 

region in which the event took place and is augmented by personnel from 

headquarters or other regions or by contractors. The augmented inspection team 

reports directly to the appropriate regional administrator. 
Appendix IV: NRC Safety Cornerstones and 

Related Inspection Procedures

Page 68 GAO-13-743 Oversight of Reactor Safety

Inspection

procedure Title Description 

Incident Investigation Team This inspection focuses on assessing a significant operational event and its causes 

within an incident investigation team, including collecting, analyzing, and 

documenting information and evidence about the event.

a

The team conducting the inspection consists of technical experts who, to the extent 

practicable, do not have, and have not had, previous significant involvement with 

licensing and inspection activities at the affected plant. The incident investigation 

team reports directly to the Executive Director for Operations and is independent of 

regional and headquarters office management. 

Source: GAO analysis of NRC information.

Note: A significant operational event is any radiological, safeguards, or other safety-related 

operational event at an NRC-licensed facility that poses an actual or a potential hazard to public 

health and safety, property, or the environment. A significant operational event also may be referred 

to as “an event” or “an incident.” 

a

An incident investigation does not have an inspection procedure, according to an NRC official. 

Instead, it is described in NRC Management Directive 8.3, NRC Incident Investigation Program. 
Appendix V: Examples of NRC Findings and 

Violations

Page 69 GAO-13-743 Oversight of Reactor Safety

This appendix provides summary information about select examples of 

NRC findings and violations. Table 18 provides a brief description, 

nuclear power plant location, and date recorded for select examples of 

NRC findings illustrating different risk significance levels. Table 19 

provides a brief description, nuclear power plant location, and date 

recorded for select examples of NRC violations illustrating different 

severity levels.

Table 18: Select Examples of NRC Findings Illustrating Different Risk Significance Levels 

Risk 

significance Description Nuclear power plant Date recorded 

Green findinga Licensee did not verify the impact that High-Energy Line Breaks in the 

turbine building could have on safety-related electrical equipment. 

Determined to be of very low safety significance because it was a 

design deficiency confirmed not to result in a loss of operability. 

Entered in corrective action plan. 

 Indian Point Nuclear 

Generating

Nov. 2012 

Licensee failed to establish and perform adequate preventive 

maintenance on a certain transformer. 

Limerick Generating 

Station

Sept. 2012 

White findingb Failure to establish and maintain emergency diesel generator 

maintenance procedures as required by regulation. 

 North Anna Power 

Station

Dec. 2011 

Flange on diesel generator failed, causing significant oil leak, and was 

required to be shut down. 

Byron Station Feb. 2011 

Yellow findingc Licensee failed to provide adequate oversight of contractor during 

maintenance of a startup transformer. Contractor replaced wires 

without necessary insulation, causing the wires to touch during 

operation, which caused a loss of off-site power to equipment. 

 Wolf Creek Generating 

Station

Jan. 2012 

Failure to maintain external flood procedures—procedures that would 

protect intake structures and auxiliary building during external floods. 

The procedure to stack sandbags at certain height over floodgates 

deemed insufficient. 

Fort Calhoun Station June 2010 

Red findingd Failure to ensure that design requirements for electrical power 

distribution system were properly implemented and maintained. Led to 

catastrophic fire that affected the plant’s ability to shut down safely. 

 Fort Calhoun Station Apr. 2012 

Failure to restore a heat removal system to operating status within the 

required time. 

Browns Ferry Nuclear 

Plant

Oct. 2010 

Source: GAO analysis of NRC information.

a

Green findings are denoted as very low safety significance; they indicate that licensee performance 

is acceptable and cornerstone objectives are fully met with nominal risk and deviation.

b

White findings are denoted as low to moderate safety significance; they indicate that licensee 

performance is acceptable, but outside the nominal risk range. Cornerstone objectives are met with 

minimal reduction in safety margin.

c

Yellow findings are denoted as substantial safety significance; they indicate a decline in licensee 

performance that is still acceptable with cornerstone objectives met, but with significant reduction in 

safety margin.

d

Appendix V: Examples of NRC Findings and 

Violations 

Red findings are denoted as high safety significance; they indicate a decline in licensee performance 

that is associated with an unacceptable loss of safety margin. Sufficient safety margin still exists to 

prevent undue risk to public health and safety.
Appendix V: Examples of NRC Findings and 

Violations

Page 70 GAO-13-743 Oversight of Reactor Safety

Table 19: Select Examples of NRC Violations Illustrating Different Severity Levels 

Severity Level Description

Nuclear

power plant Date recorded

Severity Level Ia Licensee submitted incomplete and inaccurate information related to 

completed corrective actions. Specifically, licensee did not remove 

accumulated boron deposits in certain areas. 

 Davis-Besse Nuclear 

Power Station 

Aug. 2002 

Severity Level IIb Failure to determine cause of having boric acid on reactor head and 

taking steps to preclude repetition. 

 Davis-Besse Nuclear 

Power Station 

Aug. 2002 

Licensee did not select a former employee to a competitive position 

due, in part, to the employee’s engagement in protected activities. The 

Atomic Safety and Licensing Board upheld NRC finding that licensee 

discriminated against its former employee. 

Watts Bar Nuclear 

Plant

June 2001 

Severity Level IIIc A senior reactor operator performed licensed activities while license 

was inactive and was on temporary medical hold. 

 San Onofre Nuclear 

Generating Station

Apr. 2011 

Failure to provide complete and accurate information to NRC which 

impacted a licensing decision. 

Donald C. Cook 

Nuclear Power Plant 

Oct. 2006 

Severity Level IVd Failure to submit Licensee Event Report per 10 C.F.R. 50.73(a)(2)(vii). Braidwood Station Dec. 2011 

Failure to obtain license amendment prior to implementing a proposed 

change to auxiliary feedwater system (as required by 10 C.F.R. 50.59). 

Braidwood Station Sept. 2011 

Source: GAO analysis of NRC information.

a

Severity Level I violations are violations that resulted in or could have resulted in serious safety 

consequences (e.g., violations that created the substantial potential for serious safety consequences 

or violations that involved systems failing when actually called on to prevent or mitigate a serious 

safety event).

b

Severity Level II violations are violations that resulted in or could have resulted in significant safety 

consequences (e.g., violations that created the potential for substantial safety consequences or 

violations that involved systems not being capable, for an extended period, of preventing or mitigating 

a serious safety event).

c

Severity Level III violations are violations that resulted in or could have resulted in moderate safety 

consequences (e.g., violations that created a potential for moderate safety consequences or 

violations that involved systems not being capable, for a relatively short period, of preventing or 

mitigating a serious safety event).

d

Severity Level IV violations are violations that are less serious, but are of more than minor concern, 

that resulted in no or relatively inappreciable potential safety consequences (e.g., violations that 

created the potential of more than minor safety consequences).
Appendix VI: NRC Nonescalated Findings at 

Operating Nuclear Power Plants in the United 

States

Page 71 GAO-13-743 Oversight of Reactor Safety

This appendix provides performance information about the 104 

commercial nuclear power reactors at 65 power plants for which we 

analyzed NRC data and that were operating in the United States through 

the end of calendar year 2012. Table 20 lists the number of nonescalated 

findings identified at each plant from 2000 through 2012.

Table 20: NRC Nonescalated Findings (Green Inspection Findings and Severity Level IV Violations) per Operating Commercial 

Nuclear Power Plant, Calendar Years 2000-2012 

Plant

2000

a

200120022003200420052006200720082009201020112012

Total

NRC Region I 

Beaver Valley Power Station 11 14 10 10 8 9 7 9 7 5 5 9 9 113

Calvert Cliffs Nuclear Power Plant 6 18 4 6 19 7 8 13 12 15 13 9 10 140

Hope Creek Generating Station 13 18 24 19 25 17 13 18 4 10 9 3 3 176

Indian Point Nuclear Generating 27 b 57 24 34 31 24 30 34 27 26 26 20 20 380

James A. FitzPatrick Nuclear Power 

Plant

28 12 7 8 5 8 2 8 8 10 8 6 5 115

Limerick Generating Station 8 15 11 16 8 4 4 4 8 8 5 9 10 110

Millstone Power Station 17 20 13 8 12 17 17 13 14 12 13 8 9 173

Nine Mile Point Nuclear Station 8 16 9 15 19 8 4 10 13 12 7 8 11 140

Oyster Creek Nuclear Generating 

Station

10 11 12 7 15 13 16 13 9 14 15 10 8 153

Peach Bottom Atomic Power Station 10 14 13 20 7 9 7 9 12 13 4 8 5 131

Pilgrim Nuclear Power Station 10 7 4 9 6 7 5 9 10 9 8 17 7 108

R.E. Ginna Nuclear Power Plant 7 4 11 14 9 10 12 7 10 11 5 5 6 111

Salem Nuclear Generating Station 7 17 14 26 25 25 13 10 16 6 4 11 5 179

Seabrook Station 5 10 6 10 9 8 10 6 3 1 3 7 7 85

Susquehanna Steam Electric Station 20 11 17 11 14 7 11 11 11 8 17 19 18 175

Three Mile Island Nuclear Station 7 19 8 7 14 13 9 10 7 7 9 4 12 126

Vermont Yankee Nuclear Power Plant 9 11 9 9 15 3 5 6 12 5 8 5 6 103

Total (NRC Region I) 203 274 196 229 241 189 173 190 183 172 159 158 151 2,518

NRC Region II 

Browns Ferry Nuclear Plant 5 c 6 5 4 10 8 10 19 7 14 18 15 14 135

Brunswick Steam Electric Plant 3 4 4 3 6 9 7 7 9 16 6 9 10 93

Catawba Nuclear Station 9 10 6 7 5 10 11 14 3 5 6 4 8 98

Crystal River Nuclear Generating Plant 5 5 4 5 5 6 2 7 5 6 7 3 0 60

Joseph M. Farley Nuclear Plant 5 8 5 5 4 11 6 5 6 14 11 10 6 96

Edwin I. Hatch Nuclear Plant 3 11 8 4 3 4 6 6 4 8 6 8 16 87

McGuire Nuclear Station 3 5 9 8 17 11 14 7 8 14 9 6 8 119

Appendix VI: NRC Nonescalated Findings at 

Operating Nuclear Power Plants in the 

United States
Appendix VI: NRC Nonescalated Findings at 

Operating Nuclear Power Plants in the United 

States

Page 72 GAO-13-743 Oversight of Reactor Safety

Plant

2000

a

200120022003200420052006200720082009201020112012

Total

North Anna Power Station 6 4 4 5 9 12 10 8 7 10 19 11 10 115

Oconee Nuclear Station 18 16 12 13 11 21 9 9 10 8 12 18 6 163

H.B. Robinson Steam Electric Plant 1 4 1 4 4 4 3 3 5 6 18 7 12 72

St. Lucie Plant 2 13 7 6 10 10 9 10 7 12 12 5 14 117

Sequoyah Nuclear Plant 14 11 16 9 7 4 8 8 6 8 8 11 13 123

Shearon Harris Nuclear Power Plant 8 5 5 5 5 6 4 7 6 5 11 15 8 90

Surry Nuclear Power Station 4 6 5 7 3 7 5 13 8 12 6 8 8 92

Turkey Point Nuclear Generating 4 6 2 9 13 9 8 13 8 9 9 9 7 106

Virgil C. Summer Nuclear Station 9 13 9 10 10 9 9 5 8 3 5 12 12 114

Vogtle Electric Generating Plant 3 4 8 6 8 3 3 7 5 7 6 6 9 75

Watts Bar Nuclear Plant 5 4 13 10 8 8 8 10 9 7 22 13 13 130

Total (NRC Region II) 107 135 123 120 138 152 132 158 121 164 191 170 174 1,885

NRC Region III 

Braidwood Station 5 12 12 6 5 8 9 13 14 17 22 33 22 178

Byron Station 6 14 10 17 14 20 9 13 19 10 15 22 13 182

Clinton Power Station 12 11 10 5 16 10 6 13 10 8 17 22 9 149

Davis-Besse Nuclear Power Station 11 5 16 25 52 13 14 15 8 9 13 14 6 201

Donald C. Cook Nuclear Power Plant 14 14 27 25 8 17 14 11 9 6 5 3 15 168

Dresden Nuclear Power Station 18 20 19 22 26 13 22 17 10 22 19 23 10 241

Duane Arnold Energy Center 2 1 7 22 12 20 13 21 8 13 18 11 14 162

Fermi 7 6 6 12 14 22 14 24 16 7 10 11 11 160

Kewaunee Power Station 17 12 9 17 24 25 25 39 8 23 19 20 18 256

Lasalle County Station 6 9 13 7 12 21 6 18 5 5 7 16 11 136

Monticello Nuclear Generating Plant 3 17 10 12 13 11 9 9 17 12 10 17 14 154

Palisades Nuclear Plant 4 23 13 13 10 8 28 25 27 14 17 31 13 226

Perry Nuclear Power Plant 5 7 10 19 16 61 16 16 26 26 18 20 16 256

Point Beach Nuclear Plant 11 15 10 30 25 27 25 28 34 16 21 13 21 276

Prairie Island Nuclear Generating Plant 10 5 6 5 10 16 13 10 14 27 25 24 21 186

Quad Cities Nuclear Power Station 35 7 21 15 14 11 32 18 14 16 7 15 12 217

Total (NRC Region III) 166 178 199 252 271 303 255 290 239 231 243 295 226 3,148

NRC Region IV

Arkansas Nuclear One 5 13 9 17 23 22 11 27 21 25 25 19 17 234

Callaway Plant 15 17 11 15 17 17 22 19 19 18 16 28 11 225

Columbia Generating Station 10 12 5 21 17 10 9 17 13 17 11 18 23 183

Comanche Peak Steam Electric Station 7 6 13 3 10 11 11 7 19 15 16 28 11 157

Cooper Nuclear Station 27 32 23 29 23 34 20 24 15 31 20 39 46 363
Appendix VI: NRC Nonescalated Findings at 

Operating Nuclear Power Plants in the United 

States

Page 73 GAO-13-743 Oversight of Reactor Safety

Plant

2000

a

200120022003200420052006200720082009201020112012

Total

Diablo Canyon Nuclear Power Plant 11 11 14 22 25 24 14 8 15 17 37 16 14 228

Fort Calhoun Station 11 7 18 12 15 21 18 18 19 13 23 15 29 219

Grand Gulf Nuclear Station 4 15 9 6 10 12 14 29 28 21 18 26 27 219

Palo Verde Nuclear Generating Station 4 16 3 4 49 32 39 43 22 29 14 18 26 299

River Bend Station 23 19 6 17 18 7 22 26 17 10 18 22 35 240

San Onofre Nuclear Generating Station 19 5 9 10 16 18 12 22 32 33 44 22 13 255

South Texas Project 11 14 7 15 13 14 7 10 14 10 10 17 9 151

Waterford Steam Electric Station 13 5 14 15 17 16 7 16 14 18 7 27 17 186

Wolf Creek Generating Station 6 5 3 7 9 7 15 27 43 44 38 38 24 266

Total (NRC Region IV) 166 177 144 193 262 245 221 293 291 301 297 333 302 3,225

Total (all regions) 642 764 662 794 912 889 781 931 834 868 890 956 853 10,776

Source: GAO analysis of NRC data.

Notes: This table includes data on the 104 commercial nuclear power reactors operating in the United 

States through the end of calendar year 2012. In February 2013, the owner of the Crystal River 

Nuclear Plant in Florida permanently shut down that site’s reactor; in May 2013, the owner of the 

Kewaunee Power Station in Wisconsin permanently shut down that site’s reactor; and in June 2013, 

the owner of the San Onofre Nuclear Generating Station in California permanently shut down that 

site’s two reactors. These actions reduced the number of operating commercial nuclear power 

reactors in the United States from 104 to 100.

Green inspection findings are denoted as very low safety significance; they indicate that licensee 

performance is acceptable and cornerstone objectives are fully met with nominal risk and deviation.

Severity Level IV violations are violations that are less serious, but are of more than minor concern, 

that resulted in no or relatively inappreciable potential safety consequences (e.g., violations that 

created the potential of more than minor safety consequences).

a

NRC reports these data by nuclear power plant as opposed to by individual reactor. Oftentimes, 

there are 2 or 3 reactors located at each plant. Therefore, data for all 104 reactors are included here, 

but at the plant level.

b

Although NRC tracks findings at Indian Point Units 2 and 3 separately, they are physically located at 

the same plant.

c

One of the Browns Ferry reactors (Unit 1) was shut down from March 1985 until May 2007 to 

address performance and management issues.
Appendix VII: Summary of NRC Performance 

Assessments for Nuclear Power Reactors 

Operating from 2001 through 2012

Page 74 GAO-13-743 Oversight of Reactor Safety

This appendix provides performance information about the 104 

commercial nuclear power reactors at 65 power plants for which we 

analyzed NRC data and that were operating in the United States through 

the end of calendar year 2012. Based on NRC’s assessment of licensee 

performance under the ROP, NRC places each of the licensee’s reactors 

into one of five performance categories on its action matrix, which 

corresponds to increased levels of oversight. The action matrix is NRC’s 

formal method of determining how much additional oversight—mostly in 

the form of supplemental inspections—is required on the basis of the 

number and risk significance of performance indicators and inspection 

findings. The level of oversight reported here corresponds to the highest 

level the plant received during the year, even if it was only a portion of the 

year. Thus, if a plant was placed into a different category each quarter, 

the highest category in which it was placed is reported here. (See table 

21.)

Table 21: Highest Level of Oversight Applied by NRC to Operating Commercial Nuclear Power Reactors for One or More 

Quarters of Each Calendar Year, 2001-2012

Key:

1. Action matrix column 1: Licensee response

2. Action matrix column 2: Regulatory response

3. Action matrix column 3: Degraded cornerstone

4. Action matrix column 4: Multiple/repetitive degraded cornerstone

5. Action matrix column 5: Unacceptable performance

6. Removed from action matrix: IMC 0350 process

Reactor

Highest level of oversight (action matrix

column assigned) during at least some portion of the year

200120022003200420052006200720082009201020112012

NRC Region I 

Beaver Valley Power Station, Unit 1 1 2 2 2 1 2 2 1 1 1 1 2 

Beaver Valley Power Station, Unit 2 1 2 2 2 1 2 2 1 1 1 1 2 

Calvert Cliffs Nuclear Power Plant, 

Unit 1

3 3 2 1 1 2 2 1 2 1 1 2 

Calvert Cliffs Nuclear Power Plant, 

Unit 2

1 2 2 2 2 1 1 1 2 2 2 2 

Hope Creek Generating Station, 

Unit 1

1 1 1 2 2 1 1 1 1 1 1 3 

Indian Point Nuclear Generating, 

Unit 2

4 4 3 2 2 2 1 1 1 1 1 1 

Indian Point Nuclear Generating, 

Unit 3

1 1 2 1 1 1 2 1 2 2 1 1 

Appendix VII: Summary of NRC Performance 

Assessments for Nuclear Power Reactors 

Operating from 2001 through 2012
Appendix VII: Summary of NRC Performance 

Assessments for Nuclear Power Reactors 

Operating from 2001 through 2012

Page 75 GAO-13-743 Oversight of Reactor Safety

Key:

1. Action matrix column 1: Licensee response

2. Action matrix column 2: Regulatory response

3. Action matrix column 3: Degraded cornerstone

4. Action matrix column 4: Multiple/repetitive degraded cornerstone

5. Action matrix column 5: Unacceptable performance

6. Removed from action matrix: IMC 0350 process

Reactor

Highest level of oversight (action matrix

column assigned) during at least some portion of the year

200120022003200420052006200720082009201020112012

James A. FitzPatrick Nuclear Power 

Plant

2 1 1 1 1 1 1 1 1 1 1 2 

Limerick Generating Station, Unit 1 2 2 1 1 1 1 1 1 1 1 1 1 

Limerick Generating Station, Unit 2 2 2 1 1 1 1 1 1 1 1 2 2 

Millstone Power Station, Unit 2 3 1 1 2 1 2 1 1 1 1 2 2 

Millstone Power Station, Unit 3 1 1 1 1 1 1 1 1 1 1 1 1 

Nine Mile Point Nuclear Station, 

Unit 1

2 2 2 1 1 1 2 1 1 1 1 2 

Nine Mile Point Nuclear Station, 

Unit 2

2 2 2 1 1 1 1 2 2 2 1 1 

Oyster Creek Nuclear Generating 

Station, Unit 1

2 2 1 2 3 2 2 1 2 2 1 1 

Peach Bottom Atomic Power Station, 

Unit 2

2 2 2 2 2 1 1 1 1 1 1 1 

Peach Bottom Atomic Power Station, 

Unit 3

2 2 2 1 1 1 1 1 1 1 1 1 

Pilgrim Nuclear Power Station 1 1 1 1 1 1 1 1 1 1 2 2 

R.E. Ginna Nuclear Power Plant 1 2 2 1 1 1 3 1 3 3 2 1 

Salem Nuclear Generating Station, 

Unit 1

1 2 2 2 1 1 3 1 1 1 1 3 

Salem Nuclear Generating Station, 

Unit 2

1 1 1 1 1 1 1 1 1 1 1 3 

Seabrook Station, Unit 1 2 1 2 1 1 2 1 1 2 2 1 2 

Susquehanna Steam Electric Station, 

Unit 1

2 2 1 1 1 1 1 1 1 2 3 3 

Susquehanna Steam Electric Station, 

Unit 2

2 2 1 1 1 1 1 1 1 1 1 1 

Three Mile Island Nuclear Station, 

Unit 1

2 1 1 1 2 2 1 1 1 1 1 1 

Vermont Yankee Nuclear Power 

Plant, Unit 1

1 3 1 2 2 2 2 1 1 1 1 1 
Appendix VII: Summary of NRC Performance 

Assessments for Nuclear Power Reactors 

Operating from 2001 through 2012

Page 76 GAO-13-743 Oversight of Reactor Safety

Key:

1. Action matrix column 1: Licensee response

2. Action matrix column 2: Regulatory response

3. Action matrix column 3: Degraded cornerstone

4. Action matrix column 4: Multiple/repetitive degraded cornerstone

5. Action matrix column 5: Unacceptable performance

6. Removed from action matrix: IMC 0350 process

Reactor

Highest level of oversight (action matrix

column assigned) during at least some portion of the year

200120022003200420052006200720082009201020112012

NRC Region II

Browns Ferry Nuclear Plant, Unit 1a a a a a a 3 a 2 3 4 4 4 

Browns Ferry Nuclear Plant, Unit 2 1 1 1 1 1 1 2 1 3 3 1 2 

Browns Ferry Nuclear Plant, Unit 3 1 1 1 1 1 1 1 1 3 3 1 2 

Brunswick Steam Electric Plant, 

Unit 1

1 1 1 1 1 2 3 1 2 2 2 2 

Brunswick Steam Electric Plant, 

Unit 2

1 1 1 2 2 2 2 1 2 2 2 2 

Catawba Nuclear Station, Unit 1 1 1 1 1 1 1 1 1 1 1 1 2 

Catawba Nuclear Station, Unit 2 1 1 1 1 1 1 1 1 1 1 1 1 

Crystal River Nuclear Generating 

Plant, Unit 3

2 1 1 1 2 2 1 1 1 1 2 2 

Edwin I. Hatch Nuclear Plant, Unit 1 1 1 1 1 2 2 1 1 2 1 1 1 

Edwin I. Hatch Nuclear Plant, Unit 2 1 1 1 1 2 2 2 2 3 3 1 1 

H. B. Robinson Steam Electric Plant, 

Unit 2

1 1 1 2 1 1 1 1 1 3 3 1 

Joseph M. Farley Nuclear Plant, 

Unit 1

2 1 1 1 1 2 3 3 2 2 1 2 

Joseph M. Farley Nuclear Plant, 

Unit 2

2 1 1 1 1 2 3 3 2 2 1 2 

McGuire Nuclear Station, Unit 1 1 1 1 1 1 1 1 2 2 2 1 1 

McGuire Nuclear Station, Unit 2 1 1 1 1 1 1 1 2 2 2 1 1 

North Anna Power Station, Unit 1 1 1 1 1 1 1 1 1 1 1 2 2 

North Anna Power Station, Unit 2 2 1 1 1 1 1 1 1 1 2 2 2 

Oconee Nuclear Station, Unit 1 3 4 2 3 2 3 3 2 2 3 1 1 

Oconee Nuclear Station, Unit 2 2 1 2 3 2 3 3 1 1 3 1 1 

Oconee Nuclear Station, Unit 3 2 2 2 3 2 3 3 1 1 3 1 1 

Sequoyah Nuclear Plant, Unit 1 2 2 1 2 2 1 1 1 1 1 2 2 

Sequoyah Nuclear Plant, Unit 2 2 2 1 1 1 1 1 1 1 1 1 1 
Appendix VII: Summary of NRC Performance 

Assessments for Nuclear Power Reactors 

Operating from 2001 through 2012

Page 77 GAO-13-743 Oversight of Reactor Safety

Key:

1. Action matrix column 1: Licensee response

2. Action matrix column 2: Regulatory response

3. Action matrix column 3: Degraded cornerstone

4. Action matrix column 4: Multiple/repetitive degraded cornerstone

5. Action matrix column 5: Unacceptable performance

6. Removed from action matrix: IMC 0350 process

Reactor

Highest level of oversight (action matrix

column assigned) during at least some portion of the year

200120022003200420052006200720082009201020112012

Shearon Harris Nuclear Power Plant, 

Unit 1

2 3 2 2 1 1 1 1 1 1 1 2 

St. Lucie Plant, Unit 1 1 1 1 1 1 1 1 1 3 3 1 3 

St. Lucie Plant, Unit 2 1 1 2 2 1 1 1 1 1 1 1 1 

Surry Nuclear Power Station, Unit 1 2 2 2 3 1 2 2 1 1 1 1 1 

Surry Nuclear Power Station, Unit 2 2 2 2 2 1 2 2 1 1 1 1 1 

Turkey Point Nuclear Generating, 

Unit 3

1 1 1 1 2 2 2 1 2 2 2 2 

Turkey Point Nuclear Generating, 

Unit 4

1 1 1 1 2 2 1 1 1 2 2 2 

Virgil C. Summer Nuclear Station, 

Unit 1

2 1 1 1 1 2 2 1 1 1 1 1 

Vogtle Electric Generating Plant, 

Unit 1

1 1 1 1 1 2 2 2 1 1 1 1 

Vogtle Electric Generating Plant, 

Unit 2

1 1 1 1 1 2 2 2 1 1 1 1 

Watts Bar Nuclear Plant, Unit 1 1 1 1 2 2 1 1 1 1 1 1 2 

NRC Region III

Braidwood Station, Unit 1 2 3 2 2 1 2 2 1 2 2 1 1 

Braidwood Station, Unit 2 1 1 1 1 1 2 2 1 1 1 1 2 

Byron Station, Unit 1 1 1 1 1 1 1 1 2 2 1 1 1 

Byron Station, Unit 2 1 1 1 1 1 2 2 2 2 1 2 1 

Clinton Power Station, Unit 1 2 2 1 1 1 1 2 1 1 1 1 1 

Davis-Besse Nuclear Power Station, 

Unit 1

1 b 6 6 6 6 1 1 1 2 2 1 2 

Donald C. Cook Nuclear Power Plant, 

Unit 1

6 2 2 2 1 1 3 1 1 1 1 1 

Donald C. Cook Nuclear Power Plant, 

Unit 2

6 3 3 3 2 1 3 1 1 1 1 1 

Dresden Nuclear Power Station, 

Unit 2

1 1 1 2 1 2 2 1 1 1 1 1 
Appendix VII: Summary of NRC Performance 

Assessments for Nuclear Power Reactors 

Operating from 2001 through 2012

Page 78 GAO-13-743 Oversight of Reactor Safety

Key:

1. Action matrix column 1: Licensee response

2. Action matrix column 2: Regulatory response

3. Action matrix column 3: Degraded cornerstone

4. Action matrix column 4: Multiple/repetitive degraded cornerstone

5. Action matrix column 5: Unacceptable performance

6. Removed from action matrix: IMC 0350 process

Reactor

Highest level of oversight (action matrix

column assigned) during at least some portion of the year

200120022003200420052006200720082009201020112012

Dresden Nuclear Power Station, 

Unit 3

1 2 2 2 1 1 1 1 2 2 1 1 

Duane Arnold Energy Center 1 1 1 1 1 1 2 1 2 2 1 1 

Fermi, Unit 2 2 2 2 2 2 2 1 1 1 1 1 2 

Kewaunee Power Station 3 2 2 2 3 3 3 2 2 1 1 1 

LaSalle County Station, Unit 1 1 1 1 1 1 1 1 1 1 1 1 1 

LaSalle County Station, Unit 2 2 2 1 1 1 1 1 1 1 1 1 1 

Monticello Nuclear Generating Plant, 

Unit 1

1 1 1 1 1 1 1 1 1 1 1 1 

Palisades Nuclear Plant 2 2 2 2 1 1 1 2 2 2 3 3 

Perry Nuclear Power Plant, Unit 1 1 2 3 4 4 4 4 2 1 1 3 3 

Point Beach Nuclear Plant, Unit 1 2 2 4 4 4 4 1 1 1 1 1 2 

Point Beach Nuclear Plant, Unit 2 2 2 4 4 4 4 1 1 1 1 1 2 

Prairie Island Nuclear Generating 

Plant, Unit 1

2 2 1 1 1 1 1 2 2 1 2 2 

Prairie Island Nuclear Generating 

Plant, Unit 2

2 2 1 1 1 1 1 1 2 2 1 2 

Quad Cities Nuclear Power Station, 

Unit 1

2 1 1 1 1 2 2 1 1 1 1 1 

Quad Cities Nuclear Power Station, 

Unit 2

2 1 1 1 1 1 1 1 1 1 1 1 

NRC Region IV

Arkansas Nuclear One, Unit 1 1 1 1 2 1 1 1 1 2 1 1 1 

Arkansas Nuclear One, Unit 2 1 1 1 1 1 1 1 1 1 1 1 1 

Callaway Plant 3 2 2 2 1 2 1 1 1 2 1 2 

Columbia Generating Station, Unit 2 3 3 1 1 2 2 1 1 2 2 1 3 

Comanche Peak Steam Electric 

Station, Unit 1

1 2 1 2 1 1 1 2 1 1 1 1 

Comanche Peak Steam Electric 

Station, Unit 2

1 2 1 1 1 1 1 1 1 1 1 1 
Appendix VII: Summary of NRC Performance 

Assessments for Nuclear Power Reactors 

Operating from 2001 through 2012

Page 79 GAO-13-743 Oversight of Reactor Safety

Key:

1. Action matrix column 1: Licensee response

2. Action matrix column 2: Regulatory response

3. Action matrix column 3: Degraded cornerstone

4. Action matrix column 4: Multiple/repetitive degraded cornerstone

5. Action matrix column 5: Unacceptable performance

6. Removed from action matrix: IMC 0350 process

Reactor

Highest level of oversight (action matrix

column assigned) during at least some portion of the year

200120022003200420052006200720082009201020112012

Cooper Nuclear Station 3 4 4 4 1 1 2 3 2 1 2 1 

Diablo Canyon Nuclear Power Plant, 

Unit 1

1 1 1 1 2 1 1 1 1 1 1 1 

Diablo Canyon Nuclear Power Plant, 

Unit 2

1 1 1 1 2 1 1 1 1 1 1 1 

Fort Calhoun Station, Unit 1 1 c 2 2 2 2 1 3 3 1 3 6 6 

Grand Gulf Nuclear Station, Unit 1 1 1 1 1 1 1 1 2 1 1 1 1 

Palo Verde Nuclear Generating 

Station, Unit 1

1 1 1 1 3 3 3 3 3 1 2 2 

Palo Verde Nuclear Generating 

Station, Unit 2

1 1 1 1 3 3 3 3 3 1 2 2 

Palo Verde Nuclear Generating 

Station, Unit 3

1 1 1 1 3 4 4 4 4 1 2 2 

River Bend Station, Unit 1 1 2 1 1 2 1 1 2 1 1 1 2 

San Onofre Nuclear Generating 

Station, Unit 2

1 1 2 2 2 1 1 2 2 2 1 2 

San Onofre Nuclear Generating 

Station, Unit 3

1 1 1 1 1 1 1 1 1 1 1 2 

South Texas Project, Unit 1 1 1 1 1 1 1 1 1 1 1 1 1 

South Texas Project, Unit 2 1 2 2 1 1 2 1 1 1 1 1 1 

Waterford Steam Electric Station, 

Unit 3

1 1 1 2 1 1 1 1 2 2 2 2 

Wolf Creek Generating Station, Unit 1 1 1 1 1 1 1 1 1 2 3 3 3 

Source: GAO analysis of NRC data.

Notes: In February 2013, the owner of the Crystal River Nuclear Plant in Florida permanently shut 

down that site’s reactor; in May 2013, the owner of the Kewaunee Power Station in Wisconsin 

permanently shut down that site’s reactor; and in June 2013, the owner of the San Onofre Nuclear 

Generating Station in California permanently shut down that site’s two reactors. These actions 

reduced the number of operating commercial nuclear power reactors in the United States from 104 to 

100, and they reduced the number of operating nuclear power plant sites from 65 to 62.

NRC periodically removes plants from oversight under the action matrix—due to significant 

performance or operational concerns regarding plants that are shut down—through a prescribed 

process that it refers to as the IMC 0350 process, which is named after the NRC Inspection Manual

chapter providing guidance for this level of oversight. Upon implementation of this process, the 
Appendix VII: Summary of NRC Performance 

Assessments for Nuclear Power Reactors 

Operating from 2001 through 2012

Page 80 GAO-13-743 Oversight of Reactor Safety

appropriate regional administrator establishes an oversight panel—typically comprised of officials 

from both headquarters and regional offices, including senior management—that determines the 

oversight activities necessary to authorize the restart of the plant’s reactors. See NRC Inspection 

Manual chapter 0350, “Oversight of Reactor Facilities in a Shutdown Condition due to Significant 

Performance and/or Operational Concerns.” 

a

One of the Browns Ferry reactors (Unit 1) was shut down from March 1985 until May 2007 to 

address performance and management issues.

b

NRC removed the Davis-Besse Nuclear Power Station from oversight under the action matrix and 

placed it under the IMC 0350 process from 2002 through 2005 due to an incident with the reactor’s 

vessel head that occurred in 2002.

c

NRC removed the Fort Calhoun Station from oversight under the action matrix and placed it under 

the IMC 0350 process in 2011 due to performance issues, as well as issues associated with the 

Missouri River flooding that affected the plant that year. As of July 2013, this plant has not been 

reinstated for oversight under the action matrix.
Appendix VIII: Substantive Cross-cutting 

Issues Identified by NRC at Operating Nuclear 

Power Reactors in the United States

Page 81 GAO-13-743 Oversight of Reactor Safety

This appendix provides performance information about the 104 

commercial nuclear power reactors at 65 power plants for which we 

analyzed NRC data and that were operating in the United States through 

the end of calendar year 2012. Table 22 lists the substantive cross-cutting 

issues assigned to each reactor from 2001 through 2012.

Table 22: Substantive Cross-cutting Issues Identified by NRC at Operating Commercial Nuclear Power Reactors for One or 

More Quarters of Each Calendar Year, 2001-2012 

Key:

HU: human performance

PIR: problem identification and resolution

SC: safety-conscious work environment

Reactor

200120022003200420052006200720082009201020112012

NRC Region I

Beaver Valley Power Station, Unit 1

Beaver Valley Power Station, Unit 2

Calvert Cliffs Nuclear Power Plant, Unit 

1 

PIR 

Calvert Cliffs Nuclear Power Plant, Unit 

2 

PIR 

Hope Creek Generating Station, Unit 1 PIR PIR 

SC 

PIR 

SC 

SC 

Indian Point Nuclear Generating, Unit 2 PIR 

HU

PIR 

HU

PIR PIR HU PIR 

HU

HU

Indian Point Nuclear Generating, Unit 3 HU HU

James A. FitzPatrick Nuclear Power 

Plant

PIR 

Limerick Generating Station, Unit 1

Limerick Generating Station, Unit 2

Millstone Power Station, Unit 2 PIR PIR 

Millstone Power Station, Unit 3

Nine Mile Point Nuclear Station, Unit 1 PIR PIR 

Nine Mile Point Nuclear Station, Unit 2 PIR PIR 

Oyster Creek Nuclear Generating 

Station, Unit 1

PIR HU PIR HU HU

Peach Bottom Atomic Power Station, 

Unit 2

PIR 

Peach Bottom Atomic Power Station, 

Unit 3

PIR 

Pilgrim Nuclear Power Station

Appendix VIII: Substantive Cross-cutting 

Issues Identified by NRC at Operating 

Nuclear Power Reactors in the United States
Appendix VIII: Substantive Cross-cutting 

Issues Identified by NRC at Operating Nuclear 

Power Reactors in the United States

Page 82 GAO-13-743 Oversight of Reactor Safety

Key:

HU: human performance

PIR: problem identification and resolution

SC: safety-conscious work environment

Reactor

200120022003200420052006200720082009201020112012

R.E. Ginna Nuclear Power Plant

Salem Nuclear Generating Station, 

Unit 1

PIR PIR 

SC 

PIR 

SC 

SC HU HU HU

Salem Nuclear Generating Station, 

Unit 2

PIR PIR 

SC 

PIR 

SC 

SC HU HU HU

Seabrook Station, Unit 1 PIR PIR 

Susquehanna Steam Electric Station, 

Unit 1

HU PIR PIR HU

PIR 

Susquehanna Steam Electric Station, 

Unit 2 

HU PIR PIR HU

PIR 

Three Mile Island Nuclear Station, Unit 

1 

HU PIR 

Vermont Yankee Nuclear Power Plant, 

Unit 1

NRC Region II

Browns Ferry Nuclear Plant, Unit 1a a a a a a a PIR PIR PIR PIR PIR 

Browns Ferry Nuclear Plant, Unit 2 PIR PIR PIR PIR PIR 

Browns Ferry Nuclear Plant, Unit 3 PIR PIR PIR PIR PIR 

Brunswick Steam Electric Plant, Unit 1

Brunswick Steam Electric Plant, Unit 2

Catawba Nuclear Station, Unit 1

Catawba Nuclear Station, Unit 2

Crystal River Nuclear Generating 

Plant, Unit 3

Edwin I. Hatch Nuclear Plant, Unit 1

Edwin I. Hatch Nuclear Plant, Unit 2

H. B. Robinson Steam Electric Plant, 

Unit 2

HU HU

Joseph M. Farley Nuclear Plant, Unit 1 HU

Joseph M. Farley Nuclear Plant, Unit 2 HU

McGuire Nuclear Station, Unit 1

McGuire Nuclear Station, Unit 2

North Anna Power Station, Unit 1 HU

North Anna Power Station, Unit 2 HU
Appendix VIII: Substantive Cross-cutting 

Issues Identified by NRC at Operating Nuclear 

Power Reactors in the United States

Page 83 GAO-13-743 Oversight of Reactor Safety

Key:

HU: human performance

PIR: problem identification and resolution

SC: safety-conscious work environment

Reactor

200120022003200420052006200720082009201020112012

Oconee Nuclear Station, Unit 1 HU

Oconee Nuclear Station, Unit 2 HU

Oconee Nuclear Station, Unit 3 HU

Sequoyah Nuclear Plant, Unit 1

Sequoyah Nuclear Plant, Unit 2

Shearon Harris Nuclear Power Plant, 

Unit 1

St. Lucie Plant, Unit 1

St. Lucie Plant, Unit 2

Surry Nuclear Power Station, Unit 1

Surry Nuclear Power Station, Unit 2

Turkey Point Nuclear Generating, Unit 

3 

PIR PIR PIR PIR PIR 

Turkey Point Nuclear Generating, Unit 

4 

PIR PIR PIR PIR PIR 

Virgil C. Summer Nuclear Station, Unit 

1 

Vogtle Electric Generating Plant, Unit 1

Vogtle Electric Generating Plant, Unit 2

Watts Bar Nuclear Plant, Unit 1 HU

NRC Region III

Braidwood Station, Unit 1 HU HU

Braidwood Station, Unit 2 HU HU

Byron Station, Unit 1 HU HU HU HU

Byron Station, Unit 2 HU HU HU HU

Clinton Power Station, Unit 1

Davis-Besse Nuclear Power Station, 

Unit 1

HU

Donald C. Cook Nuclear Power Plant, 

Unit 1

PIR PIR 

Donald C. Cook Nuclear Power Plant, 

Unit 2

PIR PIR 

Dresden Nuclear Power Station, Unit 2 HU HU HU

Dresden Nuclear Power Station, Unit 3 HU HU HU

Duane Arnold Energy Center HU HU
Appendix VIII: Substantive Cross-cutting 

Issues Identified by NRC at Operating Nuclear 

Power Reactors in the United States

Page 84 GAO-13-743 Oversight of Reactor Safety

Key:

HU: human performance

PIR: problem identification and resolution

SC: safety-conscious work environment

Reactor

200120022003200420052006200720082009201020112012

Fermi, Unit 2 HU PIR HU

Kewaunee Power Station PIR PIR PIR 

HU

PIR 

HU

PIR 

HU

LaSalle County Station, Unit 1 HU HU HU

LaSalle County Station, Unit 2 HU HU HU

Monticello Nuclear Generating Plant, 

Unit 1

HU

Palisades Nuclear Plant PIR 

HU

PIR 

HU

PIR PIR HU HU HU

Perry Nuclear Power Plant, Unit 1 PIR PIR 

HU

PIR 

HU

HU PIR 

HU

PIR 

HU

HU HU

Point Beach Nuclear Plant, Unit 1 PIR PIR 

HU

PIR 

HU

PIR 

HU

HU PIR 

HU

PIR 

HU

Point Beach Nuclear Plant, Unit 2 PIR PIR 

HU

PIR 

HU

PIR 

HU

HU PIR 

HU

PIR 

HU

Prairie Island Nuclear Generating 

Plant, Unit 1

HU HU HU

Prairie Island Nuclear Generating 

Plant, Unit 2

HU HU HU

Quad Cities Nuclear Power Station, 

Unit 1

HU HU HU HU

Quad Cities Nuclear Power Station, 

Unit 2

HU HU HU HU

NRC Region IV

Arkansas Nuclear One, Unit 1 PIR PIR HU HU

Arkansas Nuclear One, Unit 2 PIR PIR HU HU

Callaway Plant PIR PIR PIR HU HU PIR 

Columbia Generating Station, Unit 2 HU HU PIR HU PIR HU

Comanche Peak Steam Electric 

Station, Unit 1

Comanche Peak Steam Electric 

Station, Unit 2

Cooper Nuclear Station PIR PIR 

HU

PIR 

HU

PIR 

HU

PIR 

HU

PIR HU HU HU HU PIR 

HU
Appendix VIII: Substantive Cross-cutting 

Issues Identified by NRC at Operating Nuclear 

Power Reactors in the United States

Page 85 GAO-13-743 Oversight of Reactor Safety

Key:

HU: human performance

PIR: problem identification and resolution

SC: safety-conscious work environment

Reactor

200120022003200420052006200720082009201020112012

Diablo Canyon Nuclear Power Plant, 

Unit 1

HU PIR 

HU

PIR PIR PIR 

Diablo Canyon Nuclear Power Plant, 

Unit 2

HU PIR 

HU

PIR PIR PIR 

Fort Calhoun Station, Unit 1 HU HU

Grand Gulf Nuclear Station, Unit 1 HU

Palo Verde Nuclear Generating 

Station, Unit 1

PIR 

HU

PIR 

HU

PIR 

HU

PIR 

HU

PIR 

HU

Palo Verde Nuclear Generating 

Station, Unit 2

PIR 

HU

PIR 

HU

PIR 

HU

PIR 

HU

PIR 

HU

Palo Verde Nuclear Generating 

Station, Unit 3

PIR 

HU

PIR 

HU

PIR 

HU

PIR 

HU

PIR 

HU

River Bend Station, Unit 1 PIR PIR 

HU

San Onofre Nuclear Generating 

Station, Unit 2

PIR 

HU

PIR 

HU

PIR 

HU

HU

San Onofre Nuclear Generating 

Station, Unit 3

PIR 

HU

PIR 

HU

PIR 

HU

HU

South Texas Project, Unit 1

South Texas Project, Unit 2

Waterford Steam Electric Station, Unit 

3 

PIR 

Wolf Creek Generating Station, Unit 1 PIR 

HU

PIR 

HU

PIR PIR 

HU

PIR 

HU

Source: NRC. 

Note: This table includes data on the 104 commercial nuclear power reactors operating in the United 

States through the end of calendar year 2012. In February 2013, the owner of the Crystal River 

Nuclear Plant in Florida permanently shut down that site’s reactor; in May 2013, the owner of the 

Kewaunee Power Station in Wisconsin permanently shut down that site’s reactor; and in June 2013, 

the owner of the San Onofre Nuclear Generating Station in California permanently shut down that 

site’s two reactors. These actions reduced the number of operating commercial nuclear power 

reactors in the United States from 104 to 100.

a

One of the Browns Ferry reactors (Unit 1) was shut down from March 1985 until May 2007 to 

address performance and management issues.
Appendix IX: NRC Escalated Findings at 

Operating Nuclear Power Plants in the United 

States

Page 86 GAO-13-743 Oversight of Reactor Safety

This appendix provides performance information about the 104 

commercial nuclear power reactors at 65 power plants for which we 

analyzed NRC data and that were operating in the United States through 

the end of calendar year 2012. Table 23 lists the number of escalated 

findings identified at each plant from 2000 through 2012.

Table 23: NRC Escalated Findings (Greater-than-Green Inspection Findings and Severity Level I, II, and III Violations) per 

Operating Commercial Nuclear Power Plant, Calendar Years 2000-2012

Plant

2000

a

200120022003200420052006200720082009201020112012

Total

NRC Region I 

Beaver Valley Power Station 0 0 1 1 0 0 1 0 0 0 0 0 0 3 

Calvert Cliffs Nuclear Power Plant 0 1 2 0 1 0 1 0 0 1 1 0 0 7 

Hope Creek Generating Station 0 0 0 0 2 0 0 0 0 0 0 0 0 2 

Indian Point Nuclear Generating 0 b 2 1 0 0 1 0 0 0 0 0 0 0 4 

James A. Fitzpatrick Nuclear 

Power Plant 0 0 0 0 0 0 0 0 0 0 0 1 0 1 

Limerick Generating Station 0 3 0 0 0 0 0 0 0 0 0 1 0 4 

Millstone Power Station 1 0 0 0 0 0 0 0 0 0 0 1 0 2 

Nine Mile Point Nuclear Station 0 0 0 1 0 0 0 1 0 1 0 0 0 3 

Oyster Creek Nuclear Generating 

Station 0 0 0 0 2 1 0 0 0 0 0 0 0 3 

Peach Bottom Atomic Power 

Station 0 2 1 1 0 0 0 0 0 0 0 0 0 4 

Pilgrim Nuclear Power Station 0 0 0 0 0 1 0 0 0 0 0 1 0 2 

R.E. Ginna Nuclear Power Plant 0 0 1 0 0 0 0 0 1 2 0 0 0 4 

Salem Nuclear Generating Station 1 0 0 1 0 0 0 0 0 0 0 0 0 2 

Seabrook Station 0 1 0 0 0 0 0 0 0 1 0 0 1 3 

Susquehanna Steam Electric 

Station 1 1 1 0 0 0 0 0 0 1 1 0 0 5 

Three Mile Island Nuclear Station 0 1 0 0 0 1 0 0 0 0 0 0 0 2 

Vermont Yankee Nuclear Power 

Plant 0 0 0 0 2 0 1 0 0 0 0 0 0 3 

Total (NRC Region I) 3 11 7 4 7 4 3 1 1 6 2 4 1 54

NRC Region II

Browns Ferry Nuclear Plant 1 c 1 0 0 0 0 0 0 0 2 1 0 1 6 

Brunswick Steam Electric Plant 0 0 0 0 1 0 0 1 0 1 1 1 0 5 

Catawba Nuclear Station 0 0 0 0 0 1 0 0 0 0 0 0 1 2 

Crystal River Nuclear Generating 

Plant 0 0 0 0 0 1 0 0 0 0 0 1 0 2 

Appendix IX: NRC Escalated Findings at 

Operating Nuclear Power Plants in the 

United States
Appendix IX: NRC Escalated Findings at 

Operating Nuclear Power Plants in the United 

States

Page 87 GAO-13-743 Oversight of Reactor Safety

Plant

2000

a

200120022003200420052006200720082009201020112012

Total

Joseph M. Farley Nuclear Plant 0 0 0 0 0 0 0 2 1 1 0 0 0 4 

Edwin I. Hatch Nuclear Plant 0 0 0 0 0 1 0 0 0 2 0 0 0 3 

McGuire Nuclear Station 0 0 0 0 0 0 0 0 1 0 0 0 0 1 

North Anna Power Station 0 0 0 0 0 0 0 0 0 0 0 1 0 1 

Oconee Nuclear Station 0 2 3 1 1 0 2 0 1 0 3 1 0 14

H.B. Robinson Steam Electric 

Plant 0 0 0 0 0 0 0 0 0 0 3 0 0 3 

St Lucie Plant 0 0 0 0 0 0 0 0 0 1 0 0 0 1 

Sequoyah Nuclear Plant 0 1 0 0 1 0 0 0 0 0 0 0 0 2 

Shearon Harris Nuclear Power 

Plant 1 1 1 0 0 0 0 0 0 0 0 0 1 4 

Surry Nuclear Power Station 0 1 0 0 1 0 1 0 0 0 0 0 0 3 

Turkey Point Nuclear Generating 0 0 0 0 0 1 1 0 0 2 0 1 0 5 

Virgil C. Summer Nuclear Station 1 1 0 0 0 0 1 1 0 0 0 0 0 4 

Vogtle Electric Generating Plant 0 0 0 0 0 0 1 0 0 0 0 0 0 1 

Watts Bar Nuclear Plant 0 1 0 0 1 1 0 0 0 0 0 0 0 3 

Total (NRC Region II) 3 8 4 1 5 5 6 4 3 9 8 5 3 64

NRC Region III

Braidwood Station 0 0 1 0 0 0 1 0 0 1 0 0 0 3 

Byron Station 0 0 0 0 0 1 0 0 1 0 0 1 0 3 

Clinton Power Station 0 1 0 1 0 0 1 0 0 0 0 0 0 3 

Davis-Besse Nuclear Power 

Station 1 0 7 3 1 0 0 0 0 1 1 0 0 14

Donald C. Cook Nuclear Power 

Plant 0 0 2 0 2 6 1 0 0 0 0 0 0 11

Dresden Nuclear Power Station 0 2 0 1 0 0 0 0 0 1 0 0 0 4 

Duane Arnold Energy Center 0 0 0 0 0 0 3 0 0 1 0 0 0 4 

Fermi 0 1 0 0 0 0 0 0 0 0 0 0 0 1 

Kewaunee Power Station 1 1 1 0 1 2 0 1 1 0 0 0 1 9 

Lasalle County Station 0 0 0 0 0 1 1 0 0 0 0 0 0 2 

Monticello Nuclear Generating 

Plant 0 0 0 0 0 0 0 0 0 1 0 0 0 1 

Palisades Nuclear Plant 0 2 0 1 0 0 0 0 1 1 0 3 0 8 

Perry Nuclear Power Plant 0 1 1 3 1 0 0 0 0 0 0 1 1 8 

Point Beach Nuclear Plant 0 1 3 3 0 2 1 0 0 1 0 0 2 13

Prairie Island Nuclear Generating 

Plant 0 1 0 0 0 0 0 0 1 3 0 1 1 7 
Appendix IX: NRC Escalated Findings at 

Operating Nuclear Power Plants in the United 

States

Page 88 GAO-13-743 Oversight of Reactor Safety

Plant

2000

a

200120022003200420052006200720082009201020112012

Total

Quad Cities Nuclear Power 

Station 1 0 0 0 0 0 1 0 0 0 0 0 0 2 

Total (NRC Region III) 3 10 15 12 5 12 9 1 4 10 1 6 5 93

NRC Region IV

Arkansas Nuclear One 0 1 0 0 0 0 0 0 0 0 0 0 0 1 

Callaway Plant 3 2 1 1 0 0 0 0 0 0 0 0 0 7 

Columbia Generating Station 0 1 1 0 0 0 0 0 0 0 0 0 2 4 

Comanche Peak Steam Electric 

Station 0 1 1 0 0 0 0 0 0 0 0 0 0 2 

Cooper Nuclear Station 1 1 3 0 1 0 0 1 2 0 0 2 0 11

Diablo Canyon Nuclear Power 

Plant 0 0 0 0 0 0 0 0 0 0 0 0 0 0 

Fort Calhoun Station 0 0 1 0 0 1 1 2 0 0 1 1 1 8 

Grand Gulf Nuclear Station 0 0 0 0 0 0 0 0 0 0 0 0 0 0 

Palo Verde Nuclear Generating 

Station 0 0 0 0 2 1 2 0 0 0 0 0 0 5 

River Bend Station 0 0 2 0 0 0 0 0 0 0 0 0 0 2 

San Onofre Nuclear Generating 

Station 0 0 0 0 0 0 0 0 1 0 0 1 0 2 

South Texas Project 0 0 0 0 0 0 0 0 0 0 0 0 0 0 

Waterford Steam Electric Station 0 0 0 0 1 0 0 0 0 1 0 1 0 3 

Wolf Creek Generating Station 0 0 0 0 0 0 0 0 0 0 0 0 1 1 

Total (NRC Region IV) 4 6 9 1 4 2 3 3 3 1 1 5 4 46

Total (all regions) 13 35 35 18 21 23 21 9 11 26 12 20 13 257

Source: GAO analysis of NRC data. 

Notes: This table includes data on the 104 commercial nuclear power reactors operating in the United 

States through the end of calendar year 2012. In February 2013, the owner of the Crystal River 

Nuclear Plant in Florida permanently shut down that site’s reactor; in May 2013, the owner of the 

Kewaunee Power Station in Wisconsin permanently shut down that site’s reactor; and in June 2013, 

the owner of the San Onofre Nuclear Generating Station in California permanently shut down that 

site’s two reactors. These actions reduced the number of operating commercial nuclear power 

reactors in the United States from 104 to 100.

Greater-than-green findings refer to white, yellow, and red findings. White findings are denoted as low 

to moderate safety significance; they indicate that licensee performance is acceptable, but outside the 

nominal risk range. Cornerstone objectives are met with minimal reduction in safety margin. Yellow 

findings are denoted as substantial safety significance; they indicate a decline in licensee 

performance that is still acceptable with cornerstone objectives met, but with significant reduction in 

safety margin. Red findings are denoted as high safety significance; they indicate a decline in 

licensee performance that is associated with an unacceptable loss of safety margin. Sufficient safety 

margin still exists to prevent undue risk to public health and safety.

Severity Level I violations are violations that resulted in or could have resulted in serious safety 

consequences (e.g., violations that created the substantial potential for safety consequences or 

violations that involved systems failing when actually called on to prevent or mitigate a serious safety 

event). Severity Level II violations are violations that resulted in or could have resulted in significant 

safety consequences (e.g., violations that created the potential for substantial safety consequences 
Appendix IX: NRC Escalated Findings at 

Operating Nuclear Power Plants in the United 

States

Page 89 GAO-13-743 Oversight of Reactor Safety

or violations that involved systems not being capable, for an extended period, of preventing or 

mitigating a serious safety event). Severity Level III violations are violations that resulted in or could 

have resulted in moderate safety consequences (e.g., violations that created a potential for moderate 

safety consequences or violations that involved systems not being capable, for a relatively short 

period, of preventing or mitigating a serious safety event).

a

NRC reports these data by nuclear power plant as opposed to by individual reactor. Oftentimes, 

there are 2 or 3 reactors located at each plant. Therefore, data for all 104 reactors are included here, 

but at the plant level.

b

Although NRC tracks findings at Indian Point Units 2 and 3 separately, they are physically located at 

the same plant.

c

One of the Browns Ferry reactors (Unit 1) was shut down from March 1985 until May 2007 to 

address performance and management issues.
Appendix X: NRC Fukushima Near-Term Task 

Force Recommendations

Page 90 GAO-13-743 Oversight of Reactor Safety

NRC convened the Fukushima Near-Term Task Force to review its 

processes and regulations and determine whether its oversight could be 

informed by lessons learned from the March 2011 disaster at the 

Fukushima Daiichi Nuclear Power Plant in Japan. The task force 

concluded that a sequence of events like the Fukushima accident is 

unlikely to occur in the United States and that the continued operation of 

the nation’s commercial nuclear reactors does not pose an imminent risk 

to public health and safety. To improve NRC’s oversight, the task force 

recommended:

1. Establishing a logical, systematic, and coherent regulatory framework 

for adequate protection that appropriately balances defense-in-depth 

and risk considerations.

2. NRC require licensees to reevaluate and upgrade as necessary the 

design-basis seismic and flooding protection of structures, systems, 

and components for each operating reactor.

3. As part of the longer term review, that NRC evaluate potential 

enhancements to the capability to prevent or mitigate seismically 

induced fires and floods.

4. NRC strengthen station blackout mitigation capability at all operating 

and new reactors for design-basis and beyond-design-basis external 

events.

5. Requiring reliable hardened vent designs in boiling water reactor 

facilities with Mark I and Mark II containments.

6. As part of the longer term review, that NRC identify insights about 

hydrogen control and mitigation inside containment or in other 

buildings as additional information is revealed through further study of 

the Fukushima Dai-ichi accident.

7. Enhancing spent fuel pool makeup capability and instrumentation for 

the spent fuel pool.

8. Strengthening and integrating onsite emergency response capabilities 

such as emergency operating procedures, severe accident 

management guidelines, and extensive damage mitigation guidelines.

9. NRC require that facility emergency plans address prolonged station 

blackout and multiunit events.

10. As part of the longer term review, that NRC pursue additional 

emergency preparedness topics related to multiunit events and 

prolonged station blackout.

Appendix X: NRC Fukushima Near-Term Task 

Force Recommendations
Appendix X: NRC Fukushima Near-Term Task 

Force Recommendations

Page 91 GAO-13-743 Oversight of Reactor Safety

11. As part of the longer term review, that NRC should pursue emergency 

preparedness topics related to decisionmaking, radiation monitoring, 

and public education.

12. NRC strengthen regulatory oversight of licensee safety performance 

(i.e., the Reactor Oversight Process) by focusing more attention on 

defense-in-depth requirements consistent with the recommended 

defense-in-depth framework.
Appendix XI: Comments from the Nuclear 

Regulatory Commission

Page 92 GAO-13-743 Oversight of Reactor Safety

Appendix XI: Comments from the Nuclear 

Regulatory Commission
Appendix XI: Comments from the Nuclear 

Regulatory Commission

Page 93 GAO-13-743 Oversight of Reactor Safety
Appendix XI: Comments from the Nuclear 

Regulatory Commission

Page 94 GAO-13-743 Oversight of Reactor Safety
Appendix XII: GAO Contact and Staff 

Acknowledgments

Page 95 GAO-13-743 Oversight of Reactor Safety

Frank Rusco, (202) 512-3841 or ruscof@gao.gov

In addition to the contact named above, Ned Woodward (Assistant 

Director), John Barrett, Elizabeth Beardsley, John Delicath, R. Scott 

Fletcher, Kimberly Gianopoulos, Cindy Gilbert, Chad M. Gorman, 

Catherine Hurley, Jonathan Kucskar, Alison O’Neill, Dan Royer, Carol 

Herrnstadt Shulman, and Michelle R. Wong made key contributions to 

this report.

Appendix XII: GAO Contact and Staff 

Acknowledgments 

GAO Contact

Staff 

Acknowledgments 

(361413)
The Government Accountability Office, the audit, evaluation, and 

investigative arm of Congress, exists to support Congress in meeting its 

constitutional responsibilities and to help improve the performance and 

accountability of the federal government for the American people. GAO 

examines the use of public funds; evaluates federal programs and 

policies; and provides analyses, recommendations, and other assistance 

to help Congress make informed oversight, policy, and funding decisions. 

GAO’s commitment to good government is reflected in its core values of 

accountability, integrity, and reliability.

The fastest and easiest way to obtain copies of GAO documents at no 

cost is through GAO’s website (http://www.gao.gov). Each weekday 

afternoon, GAO posts on its website newly released reports, testimony, 

and correspondence. To have GAO e-mail you a list of newly posted 

products, go to http://www.gao.gov and select “E-mail Updates.”

The price of each GAO publication reflects GAO’s actual cost of 

production and distribution and depends on the number of pages in the 

publication and whether the publication is printed in color or black and 

white. Pricing and ordering information is posted on GAO’s website, 

http://www.gao.gov/ordering.htm. 

Place orders by calling (202) 512-6000, toll free (866) 801-7077, or 

TDD (202) 512-2537. 

Orders may be paid for using American Express, Discover Card, 

MasterCard, Visa, check, or money order. Call for additional information.

Connect with GAO on Facebook, Flickr, Twitter, and YouTube. 

Subscribe to our RSS Feeds or E-mail Updates. Listen to our Podcasts. 

Visit GAO on the web at www.gao.gov. 

Contact:

Website: http://www.gao.gov/fraudnet/fraudnet.htm

E-mail: fraudnet@gao.gov

Automated answering system: (800) 424-5454 or (202) 512-7470

Katherine Siggerud, Managing Director, siggerudk@gao.gov, (202) 512-

4400, U.S. Government Accountability Office, 441 G Street NW, Room 

7125, Washington, DC 20548

Chuck Young, Managing Director, youngc1@gao.gov, (202) 512-4800

U.S. Government Accountability Office, 441 G Street NW, Room 7149 

Washington, DC 20548

GAO’s Mission

Obtaining Copies of 

GAO Reports and 

Testimony 

Order by Phone

Connect with GAO

To Report Fraud, 

Waste, and Abuse in 

Federal Programs

Congressional 

Relations 

Public Affairs

Please Print on Recycled Paper.